[Federal Register Volume 71, Number 79 (Tuesday, April 25, 2006)]
[Notices]
[Pages 23952-23970]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-3901]
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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
[[Page 23953]]
the pendency before the Commission of a request for a hearing from any
person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 31, 2006 to April 13, 2006. The last
biweekly notice was published on April 11, 2006 (71 FR 18371).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
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
[[Page 23954]]
when the hearing is held. If the final determination is that the
amendment request involves no significant hazards consideration, the
Commission may issue the amendment and make it immediately effective,
notwithstanding the request for a hearing. Any hearing held would take
place after issuance of the amendment. If the final determination is
that the amendment request involves a significant hazards
consideration, any hearing held would take place before the issuance of
any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) e-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of amendment request: March 17, 2006.
Description of amendment request: The proposed amendment would
change the design criteria described in the Kewaunee Power Station
(KPS) Updated Safety Analysis Report (USAR). The change would add new
design criteria associated with internal flooding to the current
licensing basis for KPS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change provides clarification to the existing
functional requirements in the USAR by including specific design
criteria for analyzing internal flooding in order to verify the
capability of an SSC [structure, systems and components] to perform
its design function. The proposed change does not affect any of the
previously evaluated accidents in the KPS updated safety analysis
report (USAR). No SSCs, operating procedures, or administrative
controls that have the function of preventing or mitigating any of
these accidents are affected.
This proposed change to incorporate design criteria into the
USAR provides added administrative assurance that internal flooding
will be appropriately addressed, consistent with existing functional
requirements, and that safety related SSCs will not be affected by a
potential failure of a non-safety related SSC. The change does not
affect any accident initiators or the facility accident analysis.
Thus, the probability and the consequences of an accident remain
unchanged.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change to incorporate design criteria consistent
with existing functional requirements into the USAR does not change
the design function or operation of any safety related SSCs. The
proposed change documents design criteria in use and therefore does
not involve a physical change to the facility. The change,
therefore, does not create the possibility of a new or different
kind of accident due to credible new failure mechanisms,
malfunctions, or accident initiators not considered in the design
and licensing bases.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
This proposed change does not affect any margin of safety as
established in the Kewaunee USAR because it documents the design
criteria presently used and is consistent with the functional
requirements in the USAR. This proposed change provides added
administrative assurance that safety related SSCs will not be
affected by a potential failure of a non-safety related SSC due to a
postulated internal flooding event. The proposed change adds
criteria for the evaluation of internal flooding events that are
more detailed than the existing functional requirements in the USAR.
Therefore, the protection and subsequent availability of safety
related SSCs is maintained consistent with previously assumed
accident mitigation capabilities.
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: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Branch Chief: L. Raghavan.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of amendment request: January 12, 2006.
Description of amendment request: The proposed amendment would
correctly modify the wording in Technical Specification Surveillance
Requirement (SR) 3.6.6.3 Containment Cooling train cooling water flow
rate to accurately reflect the plant configuration. The current SR is
to verify flow to each train. The proposed revision to SR 3.6.6.3 would
verify flow to each cooler (plant configuration is two coolers per
train).
Basis for proposed no significant hazards consideration
determination:
[[Page 23955]]
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 TS change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change will revise Technical Specifications (TS)
Surveillance Requirement (SR) 3.6.6.3 containment cooling train
cooling water flow rate to accurately reflect the existing plant
configuration as described in the Updated Final Safety Analysis
Report (UFSAR) Sections 6.2, ``Containment Systems,'' and 9.4, ``Air
Conditioning, Heating, Cooling, and Ventilation Systems.'' The
revision will specify the appropriate testing requirements for
verification that each Containment Cooling System train Essential
Service Water (SX) flow rate to each cooling unit is >= 2660 gpm
[gallons per minute] and will therefore provide assurance that the
design flow rate assumed in the safety analyses will be achieved and
the Limited Conditions for Operation (LCO) will be met. This change
is in the conservative direction, i.e., verification of flow rate to
each cooling unit 3 2660 gpm is more conservative than
verification of the same flow rate to each cooling train that
consists of two cooling units. The performance of TS surveillance
testing is not a precursor to any accident previously evaluated.
Thus, the proposed change does not have any effect on the
probability of an accident previously evaluated.
The function of the Containment Cooling System in conjunction
with the Containment Spray System is to provide containment
atmosphere cooling to limit post accident pressure and temperature
in containment to less than design values. There is no change to the
design of the Containment Cooling System. Furthermore, the
surveillance testing specified in SR 3.6.6.3 will provide assurance
that the Containment Cooling System will perform as designed. Thus,
the radiological consequences of any accident previously evaluated
are not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change does not affect the control parameters
governing unit operation or the response of plant equipment to
transient conditions. The proposed change does not change or
introduce any new equipment, modes of system operation or failure
mechanisms.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
Prior to conversion to ITS [Improved Technical Specifications],
the SR equivalent to SR 3.6.6.3 required that each system of
containment cooling fans be demonstrated OPERABLE by ``verifying an
essential service water flow rate of greater than or equal to 2660
gpm to each cooler.'' During the ITS conversion, standard verbiage
for SR 3.6.6.3 was adopted; however, the specific plant design of
two Reactor Containment Fan Coolers (RCFCs) per Containment Cooling
train was inadvertently overlooked.
This proposed amendment would correctly modify the wording in
Technical Specifications (TS) Surveillance Requirement (SR) 3.6.6.3
Containment Cooling System to accurately reflect the Braidwood and
Byron existing plant design. The revision will provide the
appropriate testing requirements for verification that each
Containment Cooling System train SX cooling flow rate to each
cooling unit is >= 2660 gpm. This verification provides assurance
that the design flow rate assumed in the safety analyses will be
achieved; and, therefore the LCO will be met. The change for
verification of SX cooling flow rate from each cooling train to each
cooling unit is in the conservative direction and will revise the
existing non-conservative TS SR to be consistent with the plant
design as described in the UFSAR.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Daniel S. Collins.
FAL Energy Seabrook LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: March 23, 2006.
Description of amendment request: The proposed amendment would
revise the Seabrook Station, Unit No. 1 (Seabrook) Operating License
and Technical Specifications (TSs) to delete the license condition
requiring reporting of violations of other requirements (e.g.,
conditions listed in Section 2.C of the operating license). The change
is consistent with the notice published in the Federal Register on
November 4, 2005, as part of the consolidated line item improvement
process.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change involves the deletion of a reporting
requirement. The change does not affect plant equipment or operating
practices and, therefore, 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?
Response: No.
The proposed change is administrative in that it deletes a
reporting requirement. The change does not add new plant equipment,
change existing plant equipment, or affect the operating practices
of the facility. Therefore, the change does not create the
possibility of a new of different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change deletes a reporting requirement. The change
does not affect plant equipment or operation practices and,
therefore, does not involve a significant reduction in a margin of
safety.
Based upon the reasoning presented above, 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 Branch 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: March 23, 2006.
Description of amendment request: The proposed amendment would
revise the Seabrook Station Unit No. 1 (Seabrook) Technical
Specifications (TSs) consistent with the NRC-approved Revision 4 to
Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-449, ``Steam Generator Tube
Integrity.''
Additionally, the proposed amendment would revise Seabrook TS
Surveillance Requirement 4.4.6.2.1 to be consistent with NUREG-1431,
Revision 3, Improved Standard Technical Specifications Westinghouse
Plants.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments
adopting TSTF-449, including a model safety evaluation and model no
significant hazards consideration
[[Page 23956]]
(NSHC) determination, using the consolidated line item improvement
process. The NRC staff subsequently issued a notice of availability of
the models for referencing in license amendment applications in the
Federal Register on May 6, 2005 (70 FR 24126). The licensee affirmed
the applicability of the following NSHC determination in its
application dated March 23, 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change requires a SG [Steam Generator] Program that
includes performance criteria that will provide reasonable assurance
that the SG tubing will retain integrity over the full range of
operating conditions (including startup, operation in the power
range, hot standby, cooldown and all anticipated transients included
in the design specification). The SG performance criteria are based
on tube structural integrity, accident induced leakage, and
operational LEAKAGE.
A SGTR [steam generator tube rupture] event is one of the design
basis accidents that are analyzed as part of a plant's licensing
basis. In the analysis of a[n] SGTR event, a bounding primary to
secondary LEAKAGE rate equal to the operational LEAKAGE rate limits
in the licensing basis plus the LEAKAGE rate associated with a
double-ended rupture of a single tube is assumed.
For other design basis accidents such as MSLB [main steamline
break], rod ejection, and reactor coolant pump locked rotor the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). These analyses typically assume that
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or
increases to 1 gallon per minute as a result of accident induced
stresses. The accident induced leakage criterion introduced by the
proposed changes accounts for tubes that may leak during design
basis accidents. The accident induced leakage criterion limits this
leakage to no more than the value assumed in the accident analysis.
The SG performance criteria proposed change[s] to the TS[s] to
identify the standards against which tube integrity is to be
measured. Meeting the performance criteria provides reasonable
assurance that the SG tubing will remain capable of fulfilling its
specific safety function of maintaining reactor coolant pressure
boundary integrity throughout each operating cycle and in the
unlikely event of a design basis accident. The performance criteria
are only a part of the SG Program required by the proposed change to
the TS[s]. The program, defined by NEI [Nuclear Energy Institute]
97-06, Steam Generator Program Guidelines, includes a framework that
incorporates a balance of prevention, inspection, evaluation,
repair, and leakage monitoring. The proposed changes do not,
therefore, significantly increase the probability of an accident
previously evaluated.
The consequences of design basis accidents are, in part,
functions of the DOSE EQUIVALENT I-131 in the primary coolant and
the primary to secondary LEAKAGE rates resulting from an accident.
Therefore, limits are included in the plant technical specifications
for operational leakage and for DOSE EQUIVALENT I-131 in primary
coolant to ensure the plant is operated within its analyzed
condition. The typical analysis of the limiting design basis
accident assumes that primary to secondary leak rate after the
accident is 1 gallon per minute with no more than [500 gallons per
day or 720 gallons per day] in any one SG, and that the reactor
coolant activity levels of DOSE EQUIVALENT I-131 are at the TS
values before the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TSs and enhances the
requirements for SG inspections. The proposed change does not
adversely impact any other previously evaluated design basis
accident and is an improvement over the current TSs.
Therefore, the proposed change does not affect the consequences
of a SGTR accident and the probability of such an accident is
reduced. In addition, the proposed changes do not affect the
consequences of an MSLB, rod ejection, or a reactor coolant pump
locked rotor event, or other previously evaluated accident.
Criterion 2--The proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
The proposed performance based requirements are an improvement
over the requirements imposed by the current technical
specifications. Implementation of the proposed SG Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the SG Program will be an
enhancement of SG tube performance. Primary to secondary LEAKAGE
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in the margin of safety.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the SG Program are consistent with those in the
applicable design codes and standards and are an improvement over
the requirements in the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
Based upon the reasoning presented above, 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 Branch Chief: Darrell J. Roberts.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: March 7, 2006.
Description of amendment requests: The proposed amendments would
modify the Technical Specifications (TS) of the units to change the
reactor trip on turbine trip from the P-7 interlock to the P-8
interlock. Specifically, the amendment would effect changes in TS Table
3.3.1-1, ``Reactor Trip System Instrumentation,'' for Function 16,
``Turbine Trip.'' The purpose of the proposed amendment is to decrease
potentially unnecessary transients on the reactor and to increase plant
availability when the cause of a turbine trip is readily correctable.
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 as follows:
(1) Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
[[Page 23957]]
Response: No.
The proposed change revises the setpoint at which a reactor trip
will occur by changing the interlock at which it is enabled from the
P-7 interlock, at approximately 10 percent power, to the P-8
interlock, at less than or equal to 31 percent power. The P-7 and P-
8 interlocks are not accident initiators and the change to the
reactor trip setpoint does not create any new credible single
failure. An analysis has shown that a turbine trip without a reactor
trip at 31 percent power or below does not challenge the pressurizer
power operated relief valves (PORVs), thereby not adversely
affecting the probability of a small[-]break loss[-]of [-]coolant
accident due to a stuck open PORV. The consequences of accidents
previously evaluated are unaffected by this change because no change
to any accident mitigation scenario has resulted and there are no
additional challenges to fission product barrier integrity.
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.
No changes are being made to the plant that would introduce any
new accident causal mechanisms. The proposed change to the power
level at which a reactor trip on turbine trip is enabled does not
adversely affect previously identified accident initiators and does
not create any new accident initiators. The change does not affect
how the associated trip function operates. No new single failures or
accident scenarios are created by the proposed change and the
proposed change does not result in any event previously deemed
incredible being made credible.
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.
No safety analyses [will be] changed or modified as a result of
the proposed change in reactor trip setpoint. All margins associated
with the current safety analyses acceptance criteria are unaffected.
The current safety analyses remain binding. The safety systems
credited in the safety analyses will continue to be available to
perform their mitigation functions. The proposed change does not
affect the availability or operability of safety-related systems and
components.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
Based on the licensee's analysis, 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: James M. Petro, Jr., Esquire, One Cook
Place, Bridgman, MI 49106.
NRC Branch Chief: L. Raghavan.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: January 30, 2006.
Description of amendment request: The proposed change would revise
Cooper Nuclear Station (CNS) Technical Specification section 5.5.12,
``Primary Containment Leakage Rate Testing Program,'' to allow a one-
time extension of no more than 5 years for the Type A, Integrated
Leakage Rate Test (ILRT) interval. This revision is a one-time
exception to the 10-year frequency of the performance-based leakage
rate testing program for Type A tests as defined in Nuclear Energy
Institute (NEI) document NEI 94-01, Revision 0, ``Industry Guideline
for Implementing Performance-Based Option of 10 CFR part 50, appendix
J,'' pursuant to 10 CFR 50, appendix J, option B. The requested
exception is to allow the ILRT to be performed within 15 years from the
last ILRT, last performed on December 7, 1998.
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.
This license amendment proposes to revise the Technical
Specifications to allow for a one-time extension of the ILRT
interval from 10 years to 15 years. The containment function is
solely to mitigate the consequences of an accident. No design basis
accident is initiated by a failure of the containment leakage
mitigation function. The extension of the ILRT will not create any
adverse interactions with other systems that could result in
initiation of a design basis accident. Continued containment
integrity is also assured by the established programs for local
leakage rate testing and inservice inspections which are unaffected
by the proposed change. Therefore, the probability of occurrence of
an accident previously evaluated is not significantly increased.
The potential consequences of the proposed change have been
quantified by analyzing the changes in risk that would result from
extending the ILRT interval from 10 to 15 years. The increase in
risk in terms of person-rem per year within 50 miles resulting from
accidents was determined to be of a magnitude that NUREG-1493
indicates is imperceptible. NPPD [Nebraska Public Power District]
has also analyzed the increase in risk in terms of the frequency of
large early releases from accidents. The increase in the large early
release frequency resulting from the proposed extension was
determined to be within the guidelines published in Nuclear
Regulatory Commission (NRC) Regulatory Guide 1.174. Additionally,
the proposed change maintains defense-in-depth by preserving a
reasonable balance among prevention of core damage, prevention of
containment failure, and consequence mitigation. NPPD has determined
that the increase in conditional containment failure probability
from reducing the ILRT frequency from one test in 10 years to one
test in 15 years would be insignificant.
Therefore, the probability of occurrence or the consequences of
an accident previously analyzed are not significantly increased.
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 extension of the current interval for the ILRT does
not involve any change to the design or operation of any plant
structure, system, or component (SSC). The plant will continue to be
operated in the same manner. Since no changes to the design or
operation of the plant are being made, the proposed one-time
extension of the ILRT does not result in a new failure mode for an
accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed extension to the ILRT test interval will not result
in a change to the design or operation of any plant SSC used to shut
down the plant, initiate Emergency Core Cooling Systems, or isolate
the primary or secondary containment. Thus, the change will not
impact the ability of CNS to mitigate any accident or transient.
NUREG-1493, a generic study of the effects of extending containment
leakage testing, documented that an extension in the ILRT interval
from three per 10 years to one per 20 years resulted in an
imperceptible increase in risk to the public. NUREG-1493 generically
concluded that the design containment leakage rate contributes about
0.1 percent to the individual risk, and that the decrease in the
ILRT frequency would have a minimal effect on this risk since 95% of
the potential leakage paths are detected by Type B and Type C
testing. A risk assessment using the current CNS Probabilistic
Safety Assessment internal events model concluded that the risk
associated with this change is very small and not risk significant.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
[[Page 23958]]
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: David Terao.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: March 15, 2006.
Description of amendment request: The proposed amendment would
revise Cooper Nuclear Station (CNS) Technical Specification 5.5.12,
``Primary Containment Leakage Rate Testing Program,'' by adding two
sub-paragraphs to note exemptions from Section III.A and Section III.B
of Part 50 of Title 10 of the Code of Federal Regulations, Appendix J,
Option B. These two sub-paragraphs allow the leakage contribution from
the four main steam line penetrations, referred to as the Main Steam
Isolation Valve (MSIV) leakage, to be excluded.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
This proposed change to TS 5.5.12 does not modify existing
structures, systems or components (SSC's) of the plant, and it does
not introduce new SSC's. It does not change assumptions, methodology
or results of previously evaluated accidents in the Updated Safety
Analysis Report.
It does not change operating procedures or administrative
controls that affect the functions of SSC's. By excluding MSIV
leakage from Type A and Type B and C test results, this change will
make the CNS Primary Containment Leakage Rate Testing Program more
closely aligned with the assumptions used in associated accident
consequence analyses. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
This proposed change to TS 5.5.12.a does not modify existing
SSC's of the plant, and it does not introduce new SSC's. Thus, it
does not affect the design function or operation of SSC's involved,
and it does not introduce a new accident initiator. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Since MSIV leakage bypasses the containment and its filtration
system (Standby Gas Treatment System) during a Loss-of-Coolant
Accident (LOCA), the effects on release to the environment [are]
analyzed and specifically accounted for in the CNS dose analysis
methodology approved by Amendments 196 and 206. This proposed change
to exclude MSIV leakage from Type A and Type B and C test results
does not change dose analysis values, and thus, does not affect
actual margin in the dose analysis. Therefore, the proposed change
does not involve a significant reduction in an actual 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 C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: David Terao.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: December 29, 2005.
Description of amendment request: The proposed change would delete
Section 2.F of the Nine Mile Point, Unit 2 Facility Operating License
(FOL), NPF-69, which requires the licensee report violations of the
requirements contained in Section 2.C of this license. The NRC staff
issued a notice of opportunity for comment in the Federal Register on
August 29, 2005 (70 FR 51098), on possible amendments to delete this
reporting requirement, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on November 4,
2005 (70 FR 67202). The licensee affirmed the applicability of the
following NSHC determination in its application dated December 29,
2005.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change involves the deletion of a reporting
requirement. The change does not affect any plant equipment or
operating practices and therefore 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?
Response: No.
The proposed change is administrative in that it deletes a
reporting requirement. The change does not add new plant equipment,
change existing plant equipment, or affect the operating practices
of the facility. Therefore, the change will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change deletes a reporting requirement. The change
does not affect plant equipment or operating practices and therefore
does not involve a significant reduction in a margin of safety.
Based on the above, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Richard J. Laufer.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: March 23, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.3.4, ``Loss of Power (LOP) Diesel
Generator (DG) Start and Load Sequence Instrumentation''. The revision
modifies the section title and corrects a nonconservatism in the
degraded voltage time delay values in TS Surveillance Requirement (SR)
3.3.4.3.b.
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. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant increase in
the probability or consequences of any accident previously
evaluated.
The diesel generators (DGs) provide emergency electrical power
to the safeguard
[[Page 23959]]
buses in support of equipment required to mitigate the consequences
of design basis accidents and anticipated operational occurrences,
including an assumed loss of all offsite power. SR 3.3.4.3 verifies
that the loss of power (LOP) DG start instrumentation channels
respond to measured parameters within the necessary range and
accuracy. The proposed amendment revises the section title and
corrects nonconservative values in the allowed time delays for the
degraded voltage protection function. The revised values are more
restrictive than the previously allowed values.
Reducing the time delays for the degraded voltage function as
proposed does not significantly increase the probability of a loss
of offsite power event. The degraded voltage analysis established
both maximum time delay limits for a degraded voltage condition and
minimum time delays to prevent premature disconnection from offsite
power. The analyzed time delay limits considered prevention of
premature disconnection from offsite power such that the probability
of an unnecessary loss of offsite power is not significantly
increased.
The proposed change does not involve any hardware changes, nor
does it affect the probability of any event initiators. There will
be no change to normal plant operating parameters, accident
mitigation capabilities, or accident analysis assumptions or inputs.
Therefore, the probability or consequences of any accident
previously evaluated will not be significantly increased as a result
of the proposed change.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a new or different kind
of accident from any accident previously evaluated.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. The revised surveillance requirements are
more restrictive and will continue to assure equipment reliability
such that plant safety is maintained or will be enhanced.
Equipment important to safety will continue to operate as
designed. The changes do not result in any event previously deemed
incredible being made credible. The changes do not result in adverse
conditions or result in any increase in the challenges to safety
systems. Therefore, operation of the Point Beach Nuclear Plant in
accordance with the proposed amendment will not create the
possibility of a new or different type of accident from any accident
previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant reduction
in a margin of safety.
The diesel generators (DGs) provide emergency electrical power
to the safeguard buses in support of equipment required to mitigate
the consequences of design basis accidents and anticipated
operational occurrences, including an assumed loss of all offsite
power. SR 3.3.4.3 verifies that the loss of power (LOP) DG start
instrumentation channels respond to measured parameters within the
necessary range and accuracy. The proposed amendment corrects
nonconservative values in the allowed time delays for the degraded
voltage protection function. The revised values are more restrictive
than the previously allowed values. The proposed change to this SR
assures that design requirements of the emergency electrical power
system continue to be met.
There are no new or significant changes to the initial
conditions contributing to accident severity or consequences. The
proposed amendment will not otherwise affect the plant protective
boundaries, will not cause a release of fission products to the
public, nor will it degrade the performance of any other structures,
systems or components (SSCs) important to safety. Therefore, the
requested change will not result in 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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: L. Raghavan.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of amendment request: February 1, 2006.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) requirements for inoperable
snubbers by adding Limiting Condition for Operation (LCO) 3.0.8 for
SSES 1 and 2. This change is based on the TS Task Force (TSTF) change
traveler TSTF-372, Revision 4. A notice of availability for this TS
improvement using the consolidated line item improvement process was
published in the Federal Register on November 24, 2004, and May 4,
2005.
The Nuclear Regulatory Commission (NRC) staff issued a notice of
availability of a model no significant hazards consideration (NSHC)
determination for referencing license amendment applications in the
Federal Register on November 24, 2004 (69 FR 68412), and May 4, 2005
(70 FR 23252). The licensee affirmed the applicability of the model
NSHC determination in its application dated February 1, 2006.
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. Criterion 1--The Proposed Change Does Not Involve a
Significant Increase in the Probability or Consequences of an
Accident Previously Evaluated.
The proposed change allows a delay time for entering a supported
system technical specification (TS) when the inoperability is due
solely to an inoperable snubber if risk is assessed and managed. The
postulated seismic event requiring snubbers is a low-probability
occurrence and the overall TS system safety function would still be
available for the vast majority of anticipated challenges.
Therefore, the probability of an accident previously evaluated is
not significantly increased, if at all. The consequences of an
accident while relying on allowance provided by proposed LCO 3.0.8
are no different than the consequences of an accident while relying
on the TS required actions in effect without the allowance provided
by proposed LCO 3.0.8. Therefore the consequences of an accident
previously evaluated are not significantly affected by this change.
The addition of a requirement to assess and manage the risk
introduced by this change will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Criterion 2--The Proposed Change Does Not Create the
Possibility of a New or Different Kind of Accident From Any
Previously Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and managed, will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change will further
minimize possible concerns.
Thus, this change does not create the possibility of a new or
different kind of accident from an accident previously evaluated.
3. Criterion 3--The Proposed Change Does Not Involve a
Significant Reduction in the Margin of Safety.
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in RG 1.177. A bounding risk assessment was performed to
justify the proposed TS changes. This application of LCO 3.0.8 is
predicated upon the licensee's performance
[[Page 23960]]
of a risk assessment and the management of plant risk. The net
change to the margin of safety is insignificant. Therefore, this
change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Richard J. Laufer.
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: March 28, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specification Surveillance Requirement 3.5.1.4 by
changing the method and sample frequency for boron concentration
verification for the emergency core cooling system (ECCS) accumulators.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The ECCS Accumulators are used only to respond to an accident
and are not an accident initiator. Therefore, the probability of an
accident has not increased.
Boron concentration is controlled in the ECCS Accumulators to
prevent either excessive boron concentrations or insufficient boron
concentrations. Post-loss-of-coolant accident (LOCA) emergency
procedures directing the operator to establish simultaneous hot and
cold leg injection are based on the worst case minimum boron
precipitation time. Maintaining the maximum ECCS Accumulator boron
concentration within the upper limit ensures that the ECCS
Accumulators do not invalidate these steps. The minimum boron
requirements of 2100 (2550 after EPU [extended power uprate]) ppm
[parts per million] ppm are based on beginning-of-life reactivity
values and are selected to ensure that the reactor will remain
subcritical during the reflood stage of a large break LOCA. During a
large break LOCA, all control element assemblies are assumed not to
insert into the core, and the initial reactor shutdown is
accomplished by void formation during blowdown. Sufficient boron
concentration must be maintained in the ECCS Accumulators to prevent
a return to criticality during reflood. Level and pressure
instrumentation is provided to monitor the availability of the ECCS
Accumulators during plant operation.
The Technical Specification Surveillance Requirement (SR
3.5.1.4) verifies that the boron concentration remains within the
required range by sampling. Currently, the boron concentration in
each ECCS Accumulator is required to be verified by taking a sample
of the water in the ECCS Accumulator every 31 days on a staggered
test basis. A containment entry is required to take a sample from
each of the two ECCS Accumulators. In addition, the makeup water
source for the ECCS Accumulators is from the RWST [refueling water
storage tank], which is maintained between 2300 ppm and 2600 ppm
(2750 and 3050 after EPU) by SR 3.5.4.2, ensuring the ECCS
Accumulators are not diluted during makeup/fill evolutions. However,
the Reactor Coolant System boron concentration is lower during power
operation than the boron concentration in the ECCS Accumulators. Two
check valves in series prevent leakage from the Reactor Coolant
System into the ECCS Accumulators.
This proposed amendment would require inleakage monitoring to be
done every twelve hours in addition to taking samples from each ECCS
Accumulator every six months. Samples would continue to be taken to
verify the inleakage observations remain conservative.
The engineering analysis and risk insights combine to
demonstrate that the method of ECCS Accumulator boron concentration
verification can be changed from sampling every 31 days on a
staggered test basis to monitoring inleakage every twelve hours and
sampling each ECCS Accumulator every six months. The inleakage
monitoring is based on a calculational method that has sufficient
conservatism to predict the boron concentration of the ECCS
Accumulator as shown by sample. Therefore, the ECCS Accumulator
would remain capable of responding to an accident as described above
and the consequences of an accident previously evaluated are not
increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
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 alter the function of any
equipment, nor cause it to operate differently than it was designed
to operate. All equipment required to mitigate the consequences of
an accident would continue to operate as before. The proposed change
alters the method of verification of the ECCS Accumulator boron
concentration, but not the boron concentration requirements
themselves.
Therefore, this change does not create the possibility of a new
or different [kind] of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The inleakage monitoring done to verify the concentration of
boron in the ECCS Accumulators, is sufficiently conservative to
ensure that a decrease in boron concentration would be detected,
leading to attempts to increase the boron concentration or a need to
sample the affected ECCS Accumulator. Sampling of the ECCS
Accumulators every six months will continue to be done to ensure
that the inleakage monitoring remains conservative and
representative. If the boron concentration is maintained in the ECCS
Accumulators, the system operates as assumed in the Updated Final
Safety Analysis Report Chapter 15 analyses and the analyses
continues to meet the dose consequences acceptance criteria given in
the Updated Final Safety Analysis Report.
Therefore, this 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: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Branch Chief: Richard J. Laufer.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant (FNP), Units 1 and 2, Houston
County, Alabama; Docket Nos. 50-321 and 50-366, Edwin I. Hatch Nuclear
Plant (HNP), Units 1 and 2, Appling County, Georgia; and Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant (VEGP), Units 1 and
2, Burke County, Georgia
Date of amendment request: February 17, 2006.
Description of amendment request: The proposed amendment would add
Technical Specification (TS) Limiting Condition for Operation (LCO)
3.0.8 (and renumber existing LCO 3.0.8 to LCO 3.0.9 for VEGP) to allow
a delay time for entering a supported system TS when the inoperability
is due solely to an inoperable snubber, if risk is assessed and managed
consistent with the program in place for complying with the
requirements of 10 CFR 50.65(a)(4).
The NRC staff issued a notice of availability of a model safety
evaluation and model no significant hazards consideration (NSHC)
determination for referencing in license amendment applications in the
Federal Register on November 24, 2004 (69 FR 68412). The licensee
affirmed the applicability of the
[[Page 23961]]
model NSHC determination in its application dated February 17, 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change allows a delay time for entering a supported
system technical specification (TS) when the inoperability is due
solely to an inoperable snubber if risk is assessed and managed. The
postulated seismic event requiring snubbers is a low-probability
occurrence and the overall TS system safety function would still be
available for the vast majority of anticipated challenges.
Therefore, the probability of an accident previously evaluated is
not significantly increased, if at all. The consequences of an
accident while relying on allowance provided by proposed LCO 3.0.8
are no different than the consequences of an accident while relying
on the TS required actions in effect without the allowance provided
by proposed LCO 3.0.8. Therefore, the consequences of an accident
previously evaluated are not significantly affected by this change.
The addition of a requirement to assess and manage the risk
introduced by this change will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is
assessed and managed, will not introduce new failure modes or
effects and will not, in the absence of other unrelated failures,
lead to an accident whose consequences exceed the consequences of
accidents previously evaluated. The addition of a requirement to
assess and manage the risk introduced by this change will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in RG 1.177. A bounding risk assessment was performed to
justify the proposed TS changes. This application of LCO 3.0.8 is
predicated upon the licensee's performance of a risk assessment and
the management of plant risk. The net change to the margin of safety
is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety. Based upon the
reasoning presented above and the previous discussion of the
amendment request, the requested change does not involve a no-
significant-hazards consideration.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorneys for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201; Mr. Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037; Mr. Arthur H.
Domby, Troutman Sanders, Nations Bank Plaza, Suite 5200, 600 Peachtree
Street, NE., Atlanta, Georgia 30308-2216.
NRC Branch Chief: Evangelos C. Marinos.
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: March 29, 2006.
Description of amendment request: The amendment would revise the
Technical Specifications (TS) to adopt Nuclear Regulatory Commission
(NRC)-approved Revision 4 to Technical Specification Task Force (TSTF)
Standard Technical Specification Change Traveler, TSTF-449, ``Steam
Generator Tube Integrity.'' The proposed amendment includes changes to
the TS definition of Leakage; TS 3.4.13, ``Reactor Coolant System,
Operational Leakage''; TS 5.5.9, ``Steam Generator (SG) Tube
Surveillance Program''; and TS 5.6.10, ``Steam Generator Tube
Inspection Report''; and adds TS 3.4.17, ``Steam Generator (SG) Tube
Integrity.'' The proposed changes are necessary in order to implement
the guidance for the industry initiative on NEI (Nuclear Energy
Institute) 97-06, ``Steam Generator Program Guidelines.''
The NRC staff published a notice of opportunity for comment in the
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments
adopting TSTF-449, including a model safety evaluation and model no
significant hazards consideration (NSHC) determination, using the
consolidated line-item improvement process. The NRC staff subsequently
issued a notice of availability of the models for referencing in
license amendment applications in the Federal Register on May 6, 2005
(70 FR 24126). The licensee affirmed the applicability of the following
NSHC determination in its application dated March 29, 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The proposed change requires a SG Program that includes
performance criteria that will provide reasonable assurance that the
SG tubing will retain integrity over the full range of operating
conditions (including startup, operation in the power range, hot
standby, cooldown and all anticipated transients included in the
design specification). The SG performance criteria are based on tube
structural integrity, accident induced leakage, and operational
LEAKAGE.
A Steam Generator Tube Rupture (SGTR) event is one of the design
basis accidents that are analyzed as part of a plant's licensing
basis. In the analysis of a SGTR event, a bounding primary to
secondary LEAKAGE rate equal to the operational LEAKAGE rate limits
in the licensing basis plus the LEAKAGE rate associated with a
double-ended rupture of a single tube is assumed.
For other design basis accidents such as Main Steam Line Break
(MSLB), rod ejection, and reactor coolant pump locked rotor the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). These analyses typically assume that
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or
increases to 1 gallon per minute as a result of accident induced
stresses. The accident induced leakage criterion introduced by the
proposed changes accounts for tubes that may leak during design
basis accidents. The accident induced leakage criterion limits this
leakage to no more than the value assumed in the accident analysis.
The SG performance criteria proposed change to the TSs
identifies the standards against which tube integrity is to be
measured. Meeting the performance criteria provides reasonable
assurance that the SG tubing will remain capable of fulfilling its
specific safety function of maintaining reactor coolant pressure
boundary integrity throughout each operating cycle and in the
unlikely event of a design-basis accident. The performance criteria
are only a part of the SG Program required by the proposed change to
the TSs. The program, defined by NEI 97-06, Steam Generator Program
Guidelines, includes a framework that incorporates a balance of
prevention, inspection, evaluation, repair, and leakage monitoring.
The proposed changes do not, therefore, significantly increase the
probability of an accident previously evaluated.
The consequences of design-basis accidents are, in part,
functions of the DOSE EQUIVALENT I-131 in the primary coolant and
the primary to secondary LEAKAGE
[[Page 23962]]
rates resulting from an accident. Therefore, limits are included in
the plant technical specifications for operational leakage and for
DOSE EQUIVALENT I-131 in primary coolant to ensure the plant is
operated within its analyzed condition. The typical analysis of the
limiting design basis accident assumes that primary to secondary
leak rate after the accident is 1 gallon per minute with no more
than 500 gallons per day in any one SG, and that the reactor coolant
activity levels of DOSE EQUIVALENT I-131 are at the TS values before
the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TSs and enhances the
requirements for SG inspections. The proposed change does not
adversely impact any other previously evaluated design basis
accident and is an improvement over the current TSs.
Therefore, the proposed change does not affect the consequences
of a SGTR accident and the probability of such an accident is
reduced. In addition, the proposed change does not affect the
consequences of an MSLB, rod ejection, or a reactor coolant pump
locked rotor event, or other previously evaluated accident.
Criteria 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident From Any Previously
Evaluated.
The proposed performance-based requirements are an improvement
over the requirements imposed by the current technical
specifications. Implementation of the proposed SG Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the SG Program will be an
enhancement of SG tube performance. Primary to secondary LEAKAGE
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the SG Program are consistent with those in the
applicable design codes and standards and are an improvement over
the requirements in the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TSs.
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 Branch Chief: Evangelos C. Marinos.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of amendment request: January 6, 2006 (TS-443).
Description of amendment request: The proposed amendment involves
the activation of thermal-hydraulic stability monitoring
instrumentation and would allow for the operation of the Oscillating
Power Range Monitor (OPRM) module in the ``armed'' mode when the unit
returns to power operations. The OPRM module of the Power Range Neutron
Monitoring System is designed to provide the licensee's solution
regarding reactor stability.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No
Operating in the region of the power-to-flow map where
instabilities can occur may cause a slight, but not significant,
increase in the possibility that an instability will occur. This
slight increase is acceptable because the OPRM Upscale trip function
automatically detects and suppresses design basis thermal-hydraulic
power oscillations prior to challenging the fuel MCPR [Minimum
Critical Power Ratio] Safety Limit. Thus, the proposed changes do
not significantly increase the probability of an accident previously
evaluated.
Since the OPRM Upscale trip function precludes challenges to the
fuel MCPR Safety Limit, the proposed changes do not involve a
significant increase in the consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No
The proposed changes do not modify the basic functional
requirements of the affected equipment nor create any new system
failure modes or sequence of events that could lead to an accident.
The worst case failure of the affected equipment is failure to
perform a mitigation action. Failure of this equipment to perform a
mitigating action does not create the possibility of a new or
different kind of accident.
No new external threats or release pathways are created.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No
The proposed changes do not revise any safety margin
requirements. The OPRM Upscale trip function is designed to meet all
requirements of General Design Criteria (GDC) 10 and 12 by
automatically detecting and suppressing design basis thermal-
hydraulic power oscillations prior to challenging the fuel MCPR
Safety Limit. Thus, the new equipment improves the ability of the
equipment to automatically enforce compliance with margins of
safety.
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: February 24, 2006 (TS-06-02).
Description of amendment request: The proposed amendment would
revise the Updated Final Safety Analysis Report (UFSAR) Section 15.5
dose analysis inputs and results for the steam generator tube rupture
(SGTR) accident. This analysis is being revised for both the current
steam generators and the revised primary and secondary side
[[Page 23963]]
mass releases associated with the new replacement steam generators,
which are scheduled to be installed during the Unit 1, Cycle 7
Refueling Outage in the Fall 2006. The analysis for the current steam
generators was revised as a result of an error identified in the
computer model used to calculate the dose consequences to the Main
Control Room subsequent to an accident.
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 postulated SGTR analysis was revised to determine the
control room operator and offsite dose due to correction of computer
model input errors and for primary and secondary side mass releases
associated with the replacement steam generators. The COROD and
Control Room Emergency Ventilating System (CREVS) computer model
input errors are software issues which affect analysis results but
do not affect operation of plant systems. Consequently, correction
of these errors does not have an affect on the probability of
occurrence of an accident. The change in the primary and secondary
side mass releases associated with the replacement steam generators
results in changes to the input to the current SGTR accident
analysis. The revised analysis results in an increase the calculated
Main Control Room (MCR) SGTR doses. However, the changes in primary
and secondary side mass releases and associated release time
sequence does not increase the probability of an accident previously
evaluated.
The COROD and CREVS computer model input errors and revised
primary and secondary side mass releases associated with the
replacement steam generators will result in an increase in the
calculated MCR pre-accident iodine spike thyroid dose; however the
resulting calculated MCR dose does not exceed 10 CFR 50, Appendix A,
General Design Criteria (GDC) 19, ``Control Room,'' dose limits as
specified in NUREG-0800, ``Standard Review Plan.'' Other offsite and
MCR doses (gamma, beta, and thyroid) associated with the SGTR
accident for the current steam generators and the replacement steam
generators either remain the same, decrease slightly or increase
slightly. These changes are within the ten percent allowable
increase criteria of NEI [Nuclear Energy Institute] 96-07, Revision
1. These doses remain within a small fraction of the 10 CFR 100,
``Reactor Site Criteria,'' and 10 CFR 50 Appendix A, GDC 19 as
specified in NUREG-0800. Consequently, the changes do not involve a
significant increase in the consequences of an accident previously
evaluated.
Based on the above, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The COROD and CREVS computer model input errors are software
issues which affect analysis results but do not result in new
accident initiators since operation of plant systems and equipment
are not affected. Thus, these input changes do not create the
possibility of new or different kind of accident from those
previously evaluated. The change in the primary and secondary side
mass releases associated with the replacement steam generators
result in changes to the input to the current SGTR accident
analysis. The revised analysis results in an increase in the
calculated MCR doses. However, the changes in primary and secondary
side mass releases and associated release time sequence do not
create the possibility of a new or different kind of accident than
previously evaluated.
Based on the above, the changes will not initiate an accident
nor create any new failure mechanisms. The changes do not result in
any event previously deemed incredible being made credible. In
addition; the changes will not result in any increase in the
challenges to safety systems. Therefore, the proposed changes do not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes to the affected UFSAR tables revise the
calculation input for offsite and MCR dose values for the SGTR
accident. The MCR thyroid dose (21 [mu]Ci/gm case) for the current
steam generators and the revised mass releases associated with the
replacement steam generators exceeds the ten percent allowable
increase criteria of NEI 96-07, Revision 1. Offsite doses for the
current steam generators remain the same and then decrease slightly
for the replacement steam generators. The MCR gamma and beta doses
(21 [mu]Ci/gm case) increase slightly for the current steam
generators and then decrease slightly for the replacement steam
generators. The MCR gamma, beta and thyroid doses (0.265 [mu]Ci/gm
case) increase slightly for the current steam generators and then
decrease slightly for the revised mass releases associated with the
replacement steam generators.
The above changes in SGTR accident doses are acceptable since
the MCR doses do not exceed the requirements in 10 CFR 50, Appendix
A, GDC 19 and the whole body and thyroid doses at the exclusion area
and the lower population zone outer boundaries remain the same or
decrease relative to the UFSAR values. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902
NRC Branch Chief: Michael L. Marshall, Jr.
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
[[Page 23964]]
located in ADAMS, contact the PDR Reference staff at 1 (800) 397-4209,
(301) 415-4737 or by e-mail to [email protected].
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: August 11, 2005, as
supplemented by letters dated October 11, November 16, and December 12,
2005, and February 7, 2006.
Brief Description of amendments: The amendments revise Technical
Specification (TS) Surveillance Requirement 3.6.1.3.9 with respect to
the allowed leakage rate through each Main Steam Isolation Valve.
Date of issuance: March 2, 2006.
Effective date: March 2, 2006.
Amendment Nos.: 239 and 267.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the TS.
Date of initial notice in Federal Register: September 13, 2005 (70
FR 54087). The letters dated October 11, November 16, and December 12,
2005, and February 7, 2006, provided clarifying information that was
within the scope of the initial notice and did not change the initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 2, 2006.
No significant hazards consideration comments received: No.
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 application for amendment: April 6, 2005, as supplemented
by letters dated August 8, and December 9, 2005.
Brief description of amendment: This amendment revises Technical
Specification (TS) 6.8.4.k, ``Containment Leakage Rate Testing
Program'' and TS Surveillance Requirement 4.6.1.6.1, ``Containment
Vessel Surfaces.'' Specifically, the amendment allows a one-time
extension of Appendix J to Part 50 of Title 10 of the Code of Federal
Regulation, Type A, Containment Integrated Leak Rate Test interval from
once in 10 years to once in 15 years.
Date of issuance: March 30, 2006.
Effective date: March 30, 2006.
Amendment No.: 122.
Facility Operating License No. NPF-63: Amendment revises the TS.
Date of initial notice in Federal Register: October 11, 2005 (70 FR
59084). The supplemental letters provided clarifying information that
was within the scope of the initial notice and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 30, 2006.
No significant hazards consideration comments received: No.
Dairyland Power Cooperative, Docket No. 50-409, La Crosse Boiling Water
Reactor, Genoa, Wisconsin
Date of amendment request: December 13, 2005.
Brief description of amendment: The amendment revises Technical
Specifications to allow waste processing components or fixtures to be
handled over the Fuel Element Storage Well (FESW), limiting the weight
of such items to 50 tons (the weight of the heavy load drop found
acceptable in the cask drop analyses performed for the La Crosse
Boiling Water Reactor FESW).
Date of issuance: April 3, 2006.
Effective date: April 3, 2006.
Amendment No.: 70.
Possession Only License No. DPR-45: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 14, 2006 (71
FR 7804).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation Report, dated April 3, 2006.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: May 24, 2005.
Brief description of amendment: The amendment revised the
applicability requirements of Technical Specification 3.7.A.5.a. and
3.7.A.i. related to primary containment oxygen concentration and
drywell-to-suppression chamber differential pressure limits.
Date of issuance: April 10, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 218.
Facility Operating License No. DPR-35: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 30, 2005 (70 FR
51380).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 10, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois; Docket
Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and
2, Rock Island County, Illinois
Date of application for amendments: June 15, 2005, as supplemented
by letters dated January 26, January 31, February 22, March 3, and
March 23, 2006.
Brief description of amendments: The amendment allows a transition
to Westinghouse SVEA-96 Optima2 fuel at Dresden Nuclear Power Station
(DNPS) and Quad Cities Nuclear Power Station (QCNPS) beginning with the
QCNPS, Unit 2 refueling outage in March 2006. Specifically, the
amendment revised Technical Specifications (TSs) Section 3.1.4,
``Control Rod Scram Times,'' TS Section 4.2.1, ``Fuel Assemblies,'' and
TS Section 5.6.5, ``Core Operating limits Report (COLR),'' to support
this transition. Additionally, a new surveillance requirement was added
to verify sodium pentaborate enrichment. The core reload analyses using
the new Westinghouse analytical methods for the affected units may
result in the need for additional TS changes to support the transition
to Westinghouse SVEA-96 Optima2 fuel, such as a change to the safety
limit minimum critical power ratio.
Date of issuance: April 4, 2006.
Effective date: As of the date of issuance and shall be implemented
prior to unit startup with a reactor core containing Westinghouse SVEA-
96 Optima2 fuel.
Amendment Nos.: 220/211, 231/227.
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications and Surveillance
Requirements.
Date of initial notice in Federal Register: July 19, 2005 (70 FR
41445).
The January 26, January 31, February 22, March 3, and March 23,
2006, supplements, contained clarifying information and did not change
the NRC staff's initial proposed finding of no significant hazards
consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 4, 2006. No significant hazards
consideration comments received: No.
[[Page 23965]]
Exelon Generation Company, LLC, Docket No. 50-265, Quad Cities Nuclear
Power Station, Unit 2, Rock Island County, Illinois
Date of application for amendments: December 15, 2005, as
supplemented by letters dated February 13 and March 3, 2006.
Brief description of amendments: The amendment revised the safety
limit minimum critical power ratio values in Technical Specification
(TS) Section 2.1.1, ``Reactor Core SLs.'' Specifically, the change
required that for Quad Cities, Unit 2, the minimum critical power ratio
(MCPR) for Global Nuclear Fuel fuel shall be >= 1.09 for two
recirculation loop operation or >= 1.10 for single recirculation loop
operation. Additionally, the change required that the MCPR for
Westinghouse fuel shall be >= 1.11 for two recirculation loop operation
or >= 1.13 for single loop operation.
Date of issuance: March 31, 2006.
Effective date: As of the date of issuance and shall be implemented
prior to unit startup with a reactor core containing Westinghouse
Optima2 fuel.
Amendment No.: 226.
Facility Operating License No. DPR-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 17, 2006 (71 FR
2591).
The February 13, 2006, and March 3, 2006, supplements, contained
clarifying information and did not change the NRC staff's initial
proposed finding of no significant hazards consideration.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 31, 2006.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: November 8, 2004, as
supplemented March 31, 2005, and February 13, 2006.
Brief description of amendment: The amendment revises Technical
Specification (TS) Section 4.4.5.4 to modify the definitions of steam
generator tube ``Plugging Limit'' and ``Tube Inspection.'' The purpose
of these modifications is to define the depth of the required tube
inspections and to clarify the plugging criteria within the tubesheet
region. The amendment also modifies TS Section 4.4.5.5, ``Reports,'' to
require a Special Report of indications found in the tubesheet region
following each inspection.
Date of Issuance: April 11, 2006.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 143.
Renewed Facility Operating License No. NPF-16: Amendment revised
the TS.
Date of initial notice in Federal Register: November 24, 2004 (69
FR 68404).
The March 31, 2005, and February 13, 2006, Supplements did not
affect the original proposed no significant hazards determination, or
expand the scope of the request as noticed in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 11, 2006.
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: April 13, 2005, as supplemented by
letter dated September 29, 2005.
Brief description of amendment: The amendment incorporated several
Technical Specification Task Force (TSTF) changes to the licensee's
Technical Specifications (TSs). The specific TSTF changes that were
incorporated are:
1. TSTF-222-A, Revision 1, ``Control Rod Scram Time Testing''--This
change modifies TS Section 3.1.4, ``Control Rod Scram Times,'' to
clarify that control rod scram time testing is required only for core
cells in which work on the control rod or drive has been performed or
fuel has been moved or replaced.
2. TSTF-275-A, Revision 0, ``Clarify Requirement for EDG [emergency
diesel generator] start signal on RPV [reactor pressure vessel] Level--
Low, Low, Low during RPV cavity flood-up''--This change modifies the TS
Section 3.3.5.1, ``ECCS [emergency core cooling system]
Instrumentation,'' to clarify that the ECCS initiation instrumentation,
identified as being required in modes 4 and 5, is required to be
operable only when the associated ECCS subsystems are required to be
operable as defined in limiting condition of operation (LCO) 3.5.2,
``ECCS--Shutdown.''
3. TSTF-300-A, Revision 0, ``Eliminate DG [diesel generator] LOCA
[loss-of-coolant accident]-Start SRs [surveillance requirements] while
in S/D [shutdown] when no ECCS is Required''--This change modifies the
TS Section 3.8.2, ``AC [alternating current] Sources--Shutdown,'' to
add an additional note to the surveillance that verifies automatic
start of the emergency diesel generators and automatic load shedding
from the emergency buses, is considered to be met without the ECCS
initiation signals operable when ECCS initiation signals are not
required to be operable per Table 3.3.5.1-1, ECCS Instrumentation.
4. TSTF-225, Revision 2, ``Fuel movement with inoperable refueling
equipment interlocks''--This change modifies TS Section 3.9.1,
``Refueling Equipment Interlocks,'' to add required actions to allow
insertion of a control rod withdrawal block and verification that all
control rods are fully inserted as alternate actions to suspending in-
vessel fuel movement in the event that one or more required refueling
equipment interlocks are inoperable.
Date of issuance: March 30, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 218.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 7, 2005 (70 FR
33216).
The supplement dated September 29, 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 March 30, 2006.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: November 8, 2005, as supplemented by
letters dated March 17 and 27, 2006.
Brief description of amendment: The amendment adds limits and
controls for the spent fuel cask loading and unloading operations in
the spent fuel pool (SFP). The change modifies the technical
specifications (TSs) by adding a new Limiting Condition for Operation
(LCO) 2.8.3(6) that establishes (1) A boron concentration requirement
during cask loading operations in the SFP, and (2) a spent fuel burnup-
initial enrichment limit in the spent fuel cask to ensure subcritical
conditions are maintained during spent fuel cask loading operations in
the SFP. In addition, the change modifies TS Tables 3-4 and 3-5, and
adds a new subsection 4.3.1.3 in Design Features 4.3.1 to describe the
spent fuel cask design
[[Page 23966]]
features. In addition, editorial changes were made mostly to make the
TSs consistent with the proposed changes and to conform pagination.
Date of issuance: April 10, 2006.
Effective date: The license amendment is effective as of its date
of issuance.
Amendment No.: 239.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: December 20, 2005 (70
FR 75494).
The March 17 and 27, 2006, supplemental letters provided
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
safety evaluation dated April 10, 2006.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of application for amendments: October 5, 2005.
Brief description of amendments: The amendments change the SSES 1
and 2 Technical Specifications (TSs) 3.4.10, ``RCS [Reactor Coolant
System] Pressure and Temperature (P/T) Limits,'' by removing the valid
P/T curve limit date and replacing it with the effective full-power
years (EFPY) of radiation exposure on each of the P/T limit curves for
SSES 1 and 2. The new P/T limit will be 35.7 EFPY for SSES 1 and 30.2
EFPY for SSES 2.
Date of issuance: March 30, 2006.
Effective date: As of the date of issuance and to be implemented
within 30 days.
Amendment Nos.: 232 and 209.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 17, 2006 (71 FR
2595).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 30, 2006.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: October 5, 2005, as
supplemented on March 31, 2006.
Brief description of amendments: These amendments revise the
Technical Specifications by eliminating the requirements to submit
monthly operating reports and occupational radiation exposure reports.
Date of issuance: April 6, 2006.
Effective date: April 6, 2006.
Amendment Nos.: 233 and 210.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
153).
The supplement dated March 31, 2006, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 6, 2006.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: October 5, 2005, as
supplemented on March 31, 2006.
Brief description of amendments: These amendments revise the
Technical Specifications by eliminating the requirements associated
with hydrogen recombiners, and hydrogen and oxygen monitors.
Date of issuance: April 6, 2006.
Effective date: As of the date of issuance and to be implemented
within 60 days of the date of issuance.
Amendment Nos.: 234 and 211.
Facility Operating License Nos. NPF-14 and NPF 22: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
152).
The supplement dated March 31, 2006, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 6, 2006.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: October 11, 2005.
Brief description of amendment: The amendment revises certain 18-
month Technical Specification (TS) surveillance requirements to
eliminate the condition that testing be conducted during shutdown
conditions.
Date of issuance: April 4, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 165.
Facility Operating License No. NPF-57: This amendment revised the
TSs.
Date of initial notice in Federal Register: January 17, 2006 (71 FR
2593).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 4, 2006.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: October 11, 2005.
Brief description of amendment: The amendment removes the Technical
Specification (TS) 3.1.5 requirement for the standby liquid control
(SLC) system to be operable in Operational Condition 5 (refueling) with
any control rod withdrawn. Corresponding changes are also made to the
SLC initiation sections of TS Tables 3.3.2-1 and 4.3.2-1.
Date of issuance: April 7, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 166.
Facility Operating License No. NPF-57: This amendment revised the
TSs.
Date of initial notice in Federal Register: January 31, 2006 (71 FR
5083).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 7, 2006.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: October 11, 2005.
Brief description of amendment: The amendment changes the Technical
Specifications (TSs) to relocate the component identification of the
[[Page 23967]]
overcurrent protective devices from TS 3/4.8.4.1 and TS 3/4.8.4.5 to
the Updated Final Safety Analysis Report.
Date of issuance: April 10, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 167.
Facility Operating License No. NPF-57: The amendment revised the
TSs.
Date of initial notice in Federal Register: March 6, 2006 (71 FR
11233).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 10, 2006.
No significant hazards consideration comments received: No.
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 amendments request: September 27, 2005.
Brief Description of amendments: The amendments revise the
Technical Specifications to eliminate the power range neutron high-flux
negative rate reactor trip function.
Date of issuance: February 27, 2006.
Effective date: As of the date of issuance and shall be implemented
prior to startup following refueling outage 21 for Unit 1 and prior to
startup following refueling outage 18 for Unit 2.
Amendment Nos.: 171 and 164.
Renewed Facility Operating License Nos. NPF-2 and NPF-8: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: November 8, 2005 (70 FR
67750).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 27, 2006.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 30, 2005.
Brief description of amendments: The amendments revise Technical
Specifications to reflect incorporation of the Westinghouse Electric
Company Best Estimate Analyzer for Core Operations--Nuclear power
distribution monitoring as described in Topical Report WCAP-124-P-A,
``BEACON--Core Monitoring and Operations Support System.''
Date of issuance: March 31, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: Unit 1-175; Unit 2-163.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications and Surveillance Requirements.
Date of initial notice in Federal Register: October 11, 2005 (70 FR
59088).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 31, 2006.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 30, 2005.
Brief description of amendments: The amendments revise Technical
Specifications to reflect incorporation of the Westinghouse Electric
Company Best Estimate Analyzer for Core Operations--Nuclear power
distribution monitoring as described in Topical Report WCAP-124-P-A,
``BEACON--Core Monitoring and Operations Support System.''
Date of issuance: March 31, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment Nos.: Unit 1-175; Unit 2-163.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications and Surveillance Requirements.
Date of initial notice in Federal Register: October 11, 2005 (70 FR
59088).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 31, 2006.
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: September 1, 2005, as
supplemented by letters dated March 16 and 30, 2006.
Brief description of amendments: The amendments temporarily revise
the reactor protection system turbine trip allowable value for low trip
system pressure from greater than or equal to 43 pounds per square inch
gauge (psig) to 39.5 psig for Operating Cycle 15.
The amendments revise Technical Specification 2.2.1, Functional
Unit 17.A allowable value in Table 2.2-1 ``Reactor Trip System
Instrumentation Setpoints.''
Date of issuance: April 6, 2006.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos. 307 and 296.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the technical specifications.
Date of initial notice in Federal Register: October 25, 2005 (70 FR
61662). The supplemental letters provided clarifying information that
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 April 6, 2006.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: April 13, 2004, as supplemented by
letters dated March 18 and August 31, 2005, and January 6, 2006.
Description of amendment: The amendments revise the Technical
Specification (TS) 3.3.2, ``Engineered Safety Features Actuation System
Instrumentation, `` Function 7.b, ``Refueling Water Storage Tank
Level--Low Low'' trip setpoint, and revise the frequency of calibration
of the level transmitters from every 9 months to every 18 months.
Date of issuance: March 30, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 125 and 125.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications and Surveillance Requirements.
Date of initial notice in Federal Register: May 11, 2004 (69 FR
26193). The March 18 and August 31, 2005, and January 6, 2006,
supplemental letters provided additional information that clarified the
application, did not expand the scope of the application as originally
noticed, and did not change the staff's original proposed no
significant hazards consideration determination as published in the
Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 30, 2006.
No significant hazards consideration comments received: No.
[[Page 23968]]
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (PDR), located at One White Flint North, Public File Area 01F21,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR part 2.
Interested persons should consult a current copy of 10 CFR 2.309, which
is available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
[[Page 23969]]
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------
\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) e-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer or the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
AmerGen Energy Company, Docket No. 50-289, Three Mile Island, Unit 1,
Dauphin County, Pennsylvania
Date of amendment request: April 6, 2006.
Description of amendment request: The amendment revised Technical
Specification (TS) 3.7.2.c, ``Unit Electric Power System,'' to increase
the TS allowed outage time with one inoperable emergency diesel
generator EDG-Y-1A from 7 days to 10 days, on a one-time basis.
Date of issuance: April 8, 2006.
Effective date: As of the date of issuance and is applicable until
the emergency diesel generator EG-Y-1A is returned to operable status
or until April 12, 2006, at 21:00 hours, whichever occurs first.
Amendment No.: 258.
Facility Operating License No. DPR-50: The amendment revised the
TSs.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No. The Commission's related evaluation of the
amendment, finding of emergency circumstances, State consultation, and
final NSHC determination are contained in a safety evaluation dated
April 8, 2006.
Attorney for licensee: Assistant General Counsel, AmerGen Energy
Company, LLC 200 Exelon Way, Kennett Square, PA 19348.
NRC Branch Chief: Darrell J. Roberts.
Arizona Public Service Company, et al., Docket No. STN 50-528, Palo
Verde Nuclear Generating Station, Unit No. 1, Maricopa County, Arizona
Date of application for amendment: March 31, 2006, as supplemented
by letters dated March 31 and April 4, 2006.
Brief description of amendment: The amendment to the Updated Final
Safety Analysis Report allows the use of an operator action as a
compensatory measure to prevent exceeding the Train A shutdown cooling
(SDC) system design basis vibration limit if a Loop 2 reactor coolant
pump (RCP) should trip or have a sheared shaft during four-RCP
operation. This compensatory measure would only be used during a one-
time 12-hour period for root cause data collection in Mode 3. After the
root cause data collection is completed, a modification will be
implemented to reduce the SDC system vibration.
Date of issuance: April 6, 2006.
Effective date: April 6, 2006, and shall be implemented within 5
days of the date of issuance.
Amendment No.: Unit 1-159.
Facility Operating License No. NPF-41: The amendment revises the
Updated Final Safety Analysis Report as set forth in the application
for amendment by licensee letter dated March 31, 2006, as supplemented.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. A public notice was published in the April 3
and 4, 2006, editions of the Arizona Republic. The notice provided an
opportunity to submit comments on the Commission's proposed NSHC
determination. No comments have been received. The Commission's related
evaluation of the amendment, finding of exigent circumstances, state
consultation, and final NSHC
[[Page 23970]]
determination are contained in a safety evaluation dated April 6, 2006.
The March 31 and April 4, 2006, supplemental letters provided
additional clarifying information, did not expand the scope of the
application as originally noticed, and did not change the NRC staff's
original proposed no significant hazards consideration determination.
Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix,
Arizona 85072-2034.
NRC Branch Chief: David Terao.
Florida Power and Light, et al., Docket No. 50-389, St. Lucie Nuclear
Plant, Unit 2, St. Lucie County, Florida
Date of amendment request: February 21, 2006.
Description of amendment request: The amendment revises the
Technical Specifications (TSs) for the Containment Ventilation System
to allow additional corrective actions for inoperable containment purge
supply and exhaust valves. These corrective actions are consistent with
the Standard TSs for Combustion Engineering plants.
Date of issuance: March 17, 2006.
Effective date: March 17, 2006.
Amendment No.: 142.
Facility Operating License No. NPF-16: Amendment revises the TSs.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. 71 FR 10566 dated March 1, 2006. The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. No comments have been received. The notice also
provided an opportunity to request a hearing by May 1, 2006, but
indicated that if the Commission makes a final NSHC determination, any
such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated March 17, 2006.
Attorney for licensee: M.S. Ross, Managing Attorney, Florida Power
& Light Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Michael L. Marshall, Jr.
Southern Nuclear Operating Company, Inc., Docket No. 50-425, Vogtle
Electric Generating Plant, Unit 2, Burke County, Georgia
Date of amendment request: March 29, 2006.
Description of amendment request: The amendment revised TS 3.7.6,
``Condensate Storage Tank (CST),'' to require two CSTs to be OPERABLE
and to increase the combined safety-related minimum volume. The
amendment also revised Surveillance Requirement 3.7.6 to reflect the
additional limit for CST volume. This amendment is needed to resume
power operation at the Vogtle Electric Generating Plant, Unit 2.
Date of issuance: March 31, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 120.
Facility Operating License No. NPF-81: Amendment revises the
technical specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No. The Commission's related evaluation of the
amendment, finding of emergency circumstances, State consultation, and
final NSHC determination are contained in a safety evaluation dated
March 31, 2006.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Branch Chief: Evangelos C. Marinos.
Dated at Rockville, Maryland, this 17th day of April 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 06-3901 Filed 4-24-06; 8:45 am]
BILLING CODE 7590-01-P