[Federal Register Volume 71, Number 233 (Tuesday, December 5, 2006)]
[Notices]
[Pages 70553-70569]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-20329]
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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 to publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from November 9, 2006, to November 21, 2006. The
last biweekly notice was published on November 21, 2006 (71 FR 67391).
[[Page 70554]]
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, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of
[[Page 70555]]
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:
Rulemaking 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].
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: April 11, 2006
Description of amendment request: The proposed amendment would add
Technical Specification (TS) Limiting Condition for Operation (LCO)
3.0.8 to allow a delay time for entering a supported system TS when the
inoperability is due solely to an inoperable snubber. The proposed
changes are consistent with approval of TS Task Force (TSTF) change
TSTF-372, Revision 4, ``Addition of LCO 3.0.8, Inoperability of
Snubbers.''
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).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows a delay time for entering a supported
system 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.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of the
plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is assessed
and managed, will not introduce new failure modes or effects and will
not, in the absence of other unrelated failures, lead to an accident
whose consequences exceed the consequences of accidents previously
evaluated. The addition of a requirement to assess and manage the risk
introduced by this change will further minimize possible concerns.
Thus, this change does not create the possibility of a new or different
kind of accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic event
requiring snubbers is a low-probability occurrence and the overall TS
system safety function would still be available for the vast majority
of anticipated challenges. The risk impact of the proposed TS changes
was assessed following the three-tiered approach recommended in
Regulatory Guide 1.177. A bounding risk assessment was performed to
justify the proposed TS changes. This application of LCO 3.0.8 is
predicated upon the licensee's performance of a risk assessment and the
management of plant risk. The net change to the margin of safety is
insignificant. Therefore, this change does not involve a significant
reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Power Company LLC, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Branch Chief: Evangelos C. Marinos.
Duke Power Company LLC, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: June 5, 2006.
[[Page 70556]]
Description of amendment request: The amendments would revise the
Technical Specifications (TSs) to clarify Surveillance Requirement (SR)
3.8.1.13 and its associated Bases to state that the SR only verifies
that non-emergency diesel generator (DG) trips are bypassed. It is
based upon, and consistent with, Industry Technical Specification Task
Force (TSTF), Standard Technical Specification Traveler, TSTF-400-A,
Revision 1, ``Clarify Surveillance Requirement on Bypass of DG
Automatic Trips.''
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. Would implementation of the changes proposed in this LAR
(License Amendment Request) involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. This LAR clarifies the purpose of Surveillance Requirement
(SR) 3.8.1.13, which is to verify that non-emergency automatic
diesel generator (DG) trips are bypassed in an accident. The DG
automatic trips and their bypasses are not initiators of any
accident that has been previously evaluated. Therefore, the
probability of any of these accidents is not significantly
increased. The function of the DG in mitigating accidents is not
changed. The revised SR continues to ensure that the DG will operate
as assumed in the accident analyses. Therefore, the consequences of
any accident previously evaluated are not affected as well.
2. Would implementation of the changes proposed in this LAR
create the possibility of a new or different kind of accident from
any accident previously evaluated?
No. The changes proposed in this LAR only clarify the purpose of
SR 3.8.1.13, which is to verify that non-emergency automatic DG
trips are bypassed in an accident. The proposed change does not
involve a physical change to the plant (no new or different type of
equipment will be installed) or a change in the methods governing
normal plant operation or testing. Thus, the changes proposed in
this LAR do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Would implementation of the changes proposed in this LAR
involve a significant reduction in a margin of safety?
No. The changes proposed in this LAR only clarify the purpose of
SR 3.8.1.13, which is to verify that non-emergency automatic DG
trips are bypassed in an accident. These changes clarify the purpose
of the SR, which is to verify that the DG is capable of performing
its assumed safety function. The safety function of the DG is
unaffected, so the changes do not affect the margin of safety.
Therefore, this LAR 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: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Power Company LLC, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Branch Chief: Evangelos C. Marinos.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: April 11, 2006.
Description of amendment request: The proposed amendment would add
Technical Specification (TS) Limiting Condition for Operation (LCO)
3.0.8 to allow a delay time for entering a supported system TS when the
inoperability is due solely to an inoperable snubber. The proposed
changes are consistent with approval of TS Task Force (TSTF) Change
TSTF-372, Revision 4, ``Addition of LCO 3.0.8, Inoperability of
Snubbers.''
The NRC staff issued a Notice of Opportunity to Comment 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).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows a delay time for entering a supported
system 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.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of the
plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is assessed
and managed, will not introduce new failure modes or effects and will
not, in the absence of other unrelated failures, lead to an accident
whose consequences exceed the consequences of accidents previously
evaluated. The addition of a requirement to assess and manage the risk
introduced by this change will further minimize possible concerns.
Thus, this change does not create the possibility of a new or different
kind of accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic event
requiring snubbers is a low-probability occurrence and the overall TS
system safety function would still be available for the vast majority
of anticipated challenges. The risk impact of the proposed TS changes
was assessed following the three-tiered approach recommended in
Regulatory Guide 1.177. A bounding risk assessment was performed to
justify the proposed TS changes. This application of LCO 3.0.8 is
predicated upon the licensee's performance of a risk assessment and the
management of plant risk. The net change to the margin of safety is
insignificant. Therefore, this change does not involve a significant
reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
[[Page 70557]]
Attorney for licensee: Ms. Lisa F. Vaughn, Duke Power Company LLC,
422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Branch Chief: Evangelos C. Marinos.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: April 11, 2006.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TSs) related to steam generator
(SG) tube integrity. The changes are consistent with the consolidated
line-item improvement process (CLIIP), Nuclear Regulatory Commission-
approved Revision 4 to Technical Specification Task Force (TSTF)
Standard TS Change Traveler, TSTF-449, ``Steam Generator Tube
Integrity.''
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 MSLB, rod ejection, and
reactor coolant pump locked rotor the tubes are assumed to retain their
structural integrity (i.e., they are assumed not to rupture). These
analyses typically assume that primary to secondary LEAKAGE for all SGs
is 1 gallon per minute or increases to 1 gallon per minute as a result
of accident induced stresses. The accident induced leakage criterion
introduced by the proposed changes accounts for tubes that may leak
during design basis accidents. The accident induced leakage criterion
limits this leakage to no more than the value assumed in the accident
analysis.
The SG performance criteria proposed change to the TS identify the
standards against which tube integrity is to be measured. Meeting the
performance criteria provides reasonable assurance that the SG tubing
will remain capable of fulfilling its specific safety function of
maintaining reactor coolant pressure boundary integrity throughout each
operating cycle and in the unlikely event of a design basis accident.
The performance criteria are only a part of the SG Program required by
the proposed change to the TS. The program, defined by NEI 97-06, Steam
Generator Program Guidelines, includes a framework that incorporates a
balance of prevention, inspection, evaluation, repair, and leakage
monitoring. The proposed changes do not, therefore, significantly
increase the probability of an accident previously evaluated.
The consequences of design basis accidents are, in part, functions
of the DOSE EQUIVALENT 1-131 in the primary coolant and the primary to
secondary LEAKAGE rates resulting from an accident. Therefore, limits
are included in the plant technical specifications for operational
leakage and for DOSE EQUIVALENT 1-131 in primary coolant to ensure the
plant is operated within its analyzed condition. The typical analysis
of the limiting design basis accident assumes that primary to secondary
leak rate after the accident is 0.27 gallons per minute with no more
than 135 gallons per day in any one SG, and that the reactor coolant
activity levels of DOSE EQUIVALENT 1-131 are at the TS values before
the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TSs and enhances the requirements
for SG inspections. The proposed change does not adversely impact any
other previously evaluated design basis accident and is an improvement
over the current TSs.
Therefore, the proposed change does not affect the consequences of
a SGTR accident and the probability of such an accident is reduced. In
addition, the proposed changes do not affect the consequences of an
MSLB (main steamline break), 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.
[[Page 70558]]
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 and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Duke Power Company LLC,
422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Branch Chief: Evangelos C. Marinos.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: June 5, 2006.
Description of amendment request: The amendments would revise the
Technical Specifications (TSs) to clarify Surveillance Requirement (SR)
3.8.1.13 and its associated Bases to state that the SR only verifies
that non-emergency diesel generator (DG) trips are bypassed. It is
based upon, and consistent with, Industry Technical Specification Task
Force (TSTF), Standard Technical Specification Traveler, TSTF-400-A,
Revision 1, ``Clarify Surveillance Requirement on Bypass of DG
Automatic Trips.''
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. Would implementation of the changes proposed in this LAR
(License Amendment Request) involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. This LAR clarifies the purpose of Surveillance Requirement
(SR) 3.8.1.13, which is to verify that non-emergency automatic
diesel generator (DG) trips are bypassed in an accident. The DG
automatic trips and their bypasses are not initiators of any
accident that has been previously evaluated. Therefore, the
probability of any of these accidents is not significantly
increased. The function of the DG in mitigating accidents is not
changed. The revised SR continues to ensure that the DG will operate
as assumed in the accident analyses. Therefore, the consequences of
any accident previously evaluated are not affected as well.
2. Would implementation of the changes proposed in this LAR
create the possibility of a new or different kind of accident from
any accident previously evaluated?
No. The changes proposed in this LAR only clarify the purpose of
SR 3.8.1.13, which is to verify that non-emergency automatic DG
trips are bypassed in an accident. The proposed change does not
involve a physical change to the plant (no new or different type of
equipment will be installed) or a change in the methods governing
normal plant operation or testing. Thus, the changes proposed in
this LAR do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Would implementation of the changes proposed in this LAR
involve a significant reduction in a margin of safety?
No. The changes proposed in this LAR only clarify the purpose of
SR 3.8.1.13, which is to verify that non-emergency automatic DG
trips are bypassed in an accident. These changes clarify the purpose
of the SR, which is to verify that the DG is capable of performing
its assumed safety function. The safety function of the DG is
unaffected, so the changes do not affect the margin of safety.
Therefore, this LAR 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: Ms. Lisa F. Vaughn, Duke Power Company LLC,
422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Branch Chief: Evangelos C. Marinos.
Duke Power Company LLC, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: July 31, 2006.
Description of amendment request: The proposed amendments would
revise Technical Specification Section 3.6.3, ``Containment Isolation
Valves,'' and its associated Bases, by removing the allowance to open
the upper containment purge isolation valves in the applicable modes
consistent with the lower containment purge isolation valves.
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 this LAR [License Amendment Request] involve a
significant increase in the probability or consequences of an
accident previously evaluated?
No. The Containment Purge System is not capable of initiating
any accident by itself so there will be no increase in the
probability of an accident. Since these containment isolation valves
will be maintained in the sealed closed position, there can be no
increase in the consequences of an accident. The design and
operation of the Containment Purge System is not being modified by
this LAR. Therefore, approval and implementation of this LAR will
have no effect on accident probabilities or consequences.
2. Does this LAR create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. This LAR does not involve any physical changes to the
Containment Purge System so no new or different accident causal
mechanisms will be generated. Also, no changes are being made to the
way in which the Containment Purge System is operated. Some
surveillance tests will no longer be performed but these tests are
no longer necessary since the affected components remain in their
safe, design basis position. Consequently, plant accident analyses
will not be affected by this LAR.
3. Does this LAR involve a significant reduction in a margin of
safety?
No. Margin of safety is related to the confidence in the ability
of the fission product barriers to perform their design functions
during and following accident conditions. These barriers include the
fuel cladding, the reactor coolant system, and the containment
system. The performance of these barriers will not be affected by
the proposed changes. The containment isolation valves in the
Containment Purge System will continue to perform their design basis
function after this LAR is implemented.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Duke Power Company LLC,
422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Branch Chief: Evangelos C. Marinos.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: November 1, 2006.
Description of amendment request: The proposed amendment would
modify technical specification (TS) requirements for inoperable
snubbers by adding Limiting Condition of Operation (LCO) 3.0.8.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on November 24, 2004 (69 FR 68412), on possible
amendments to revise the plant-specific TS to allow a
[[Page 70559]]
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 that is in place for complying with the
requirements of 10 CFR 50.65(a)(4). LCO 3.0.8 was proposed to be added
to an individual TS providing this allowance, including a model safety
evaluation and model no significant hazards consideration (NSHC)
determination, using the consolidated line-item improvement process.
The NRC staff subsequently issued a notice of availability of the
models for referencing in license amendment applications in the Federal
Register on May 4, 2005 (70 FR 23252). The licensee affirmed the
applicability of the model NSHC determination in its application dated
November 1, 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 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.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed change does not involve a physical alteration of the
plant (no new or different type of equipment will be installed).
Allowing delay times for entering supported system TS when
inoperability is due solely to inoperable snubbers, if risk is assessed
and managed, will not introduce new failure modes or effects and will
not, in the absence of other unrelated failures, lead to an accident
whose consequences exceed the consequences of accidents previously
evaluated. The addition of a requirement to assess and manage the risk
introduced by this change will further minimize possible concerns.
Thus, this change does not create the possibility of a new or different
kind of accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic event
requiring snubbers is a low-probability occurrence and the overall TS
system safety function would still be available for the vast majority
of anticipated challenges. The risk impact of the proposed TS changes
was assessed following the three-tiered approach recommended in RG
[Regulatory Guide] 1.177. A bounding risk assessment was performed to
justify the proposed TS changes. This application of LCO 3.0.8 is
predicated upon the licensee's performance of a risk assessment and the
management of plant risk. The net change to the margin of safety is
insignificant. Therefore, this change does not involve a significant
reduction in a margin of safety.
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.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: David Terao.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: November 1, 2006.
Description of amendment request: The proposed change will revise
the Grand Gulf Nuclear Station (GGNS), Unit 1, Technical Specification
(TS) Surveillance Requirement 3.3.1.1.7 for the surveillance interval
of the local power range monitor (LPRM) calibrations from 1,000
megawatt-days/ton (MWD/T) (approximately every 36 days) to 2,000 MWD/T
(approximately every 72 days).
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 extended surveillance interval continues to ensure that the
LPRM detectors are adequately calibrated to provide an accurate
indication of core power distribution and local power changes. The
change will not alter the basic operation of any process variables,
structures, systems, or components as described in the safety
analyses, and no new equipment is introduced. Hence, the probability
of accidents previously evaluated is unchanged.
The thermal limits established by safety analysis calculations
ensure that reactor core operation is maintained within fuel design
limits during any Anticipated Operational Occurrence (AOO). The
analytical methods and assumptions used in evaluating these
transients and establishing the thermal limits assure adequate
margins to fuel design limits are maintained. These methods account
for various calculation uncertainties including radial bundle power
uncertainty which can be affected by LPRM accuracy. Extending the
LPRM calibration interval does not impact the existing uncertainties
assumed in the GGNS safety analyses. Plant specific evaluation of
LPRM sensitivity to exposure has determined that the extended
calibration interval does not affect the radial bundle power
distribution uncertainty value currently used in the safety
analysis. Hence the safety analysis calculations and the associated
thermal limits are not affected by the extended LPRM calibration
interval and the consequences of an accident previously evaluated
are not changed.
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 TS amendment will not change the design function,
reliability, performance, or operation of any plant systems,
components, or structures. It does not create the possibility of a
new failure mechanism, malfunction, or accident
[[Page 70560]]
initiators not considered in the design and licensing bases. Plant
operation will continue to be within the core operating limits that
are established using NRC approved methods that are applicable to
the GGNS design and the GGNS fuel.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The thermal limits established by safety analysis calculations
ensure that reactor core operation is maintained within fuel design
limits during any Anticipated Operational Occurrence (AOO). The
analytical methods and assumptions used in evaluating these
transients and establishing the thermal limits assure adequate
margins to fuel design limits are maintained. These methods account
for various calculation uncertainties including radial bundle power
uncertainty which can be affected by LPRM accuracy. Extending the
LPRM calibration interval does not impact the existing uncertainties
assumed in the GGNS safety analyses. Plant specific evaluation of
LPRM sensitivity to exposure has determined that the extended
calibration interval does not affect the radial bundle power
distribution uncertainty value currently used in the safety
analyses. The thermal limits determined by NRC approved analytical
methods will continue to provide adequate margin to fuel design
limits.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213
NRC Branch Chief: David Terao
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: March 1, 2006
Description of amendment request: The proposed amendment would
modify the Special Operations Limiting Condition for Operation (LCO)
3.10.1, ``System Leakage and Hydrostatic Testing Operation,'' allowance
for operation with the average reactor coolant temperature greater than
212 [deg]F while considering operational conditions to be in MODE 4, to
include operations where temperature exceeds 212 [deg]F as a
consequence of maintaining reactor pressure for a system leakage or
hydrostatic test, or as a consequence of maintaining reactor pressure
for control rod scram time testing initiated in conjunction with a
system leakage or hydrostatic test. This change would allow more
efficient testing during a refueling outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Technical Specifications currently allow for operation at >212
[deg]F while imposing MODE 4 requirements in addition to the
secondary containment requirements required to be met. Extending the
activities that can apply this allowance will not adversely impact
the probability or consequences of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Technical Specifications currently allow for operation at >212
[deg]F while imposing MODE 4 requirements in addition to the
secondary containment requirements required to be met. No new
operational conditions beyond those currently allowed by LCO 3.10.1
are introduced. The extended allowances would result from operations
that commence at reduced temperatures, but approach the normal MODE
4 limit of 212 [deg]F prior to completion of the inspections or
testing. The changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the methods governing normal plant
operation. In addition, the changes do not impose any new or
different requirements or eliminate any existing requirements. The
changes do not alter assumptions made in the safety analysis. The
proposed changes are consistent with the safety analysis assumptions
and current plant operating practice.
Therefore, the 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.
Technical Specifications currently allow for operation at >212
[deg]F while imposing MODE 4 requirements in addition to the
secondary containment requirements required to be met. Extending the
activities that can apply this allowance will not adversely impact
any margin of safety. Allowing completion of inspections and testing
and supporting completion of scram time testing initiated in
conjunction with a system leakage or hydrostatic test prior to power
operation, results in enhanced safe operations by eliminating
unnecessary maneuvers to control reactor temperature and pressure.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. R. E. Helfrich, Florida Power & Light
Company, P. O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: L. Raghavan.
GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear
Station, Unit 2, Dauphin County, Pennsylvania
Date of amendment request: October 10, 2006.
Description of amendment requests: The amendment application
proposes a revision to the Technical Specification Surveillance
Requirement 4.1.1.3 to extend the containment airlock surveillance
frequency from once per year to once every five years.
Basis for proposed no significant hazards consideration
determination:
As required by 10 CFR 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration, which is
presented below:
(1) Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
No.
The proposed change does not introduce any new degradation or
failure mechanism. The failure mechanism in this case would be a
failure of an airlock door to open, thus no new release path to the
environment is created. As no release path is created, there is not
the possibility of a significant increase in the probability or
consequences of an accident.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No.
The proposed change does not introduce any new degradation or
failure mechanism.
The failure mechanism in this case would be a failure of an
airlock door to open, thus no new release path to the environment is
created. As no release path is created, there is not the possibility
of a new or different kind of accident from any accident previously
evaluated being created.
[[Page 70561]]
(3) Does the proposed change involve a significant reduction in
a margin of safety? No.
The proposed change does not introduce any new degradation or
failure mechanism. The failure mechanism in this case would be a
failure of an airlock door to open, thus no new release path to the
environment is created. Thus, 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.
NRC Branch Chief: Claudia Craig.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: October 17, 2006.
Description of amendment request: The proposed amendment would
revise the Cooper Nuclear Station (CNS) Technical Specifications (TS)
4.3.1.1.c by adding a new nominal center-to-center distance between
fuel assemblies for the new storage racks, and would revise TS 4.3.3 by
increasing the capacity of the spent fuel storage pool from 2366
assemblies to 2651 assemblies.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The probability of a seismic event, and the resulting loss of
spent fuel pool cooling flow, is not influenced by the proposed
changes. In addition, the probability of an accidental fuel assembly
drop or misloading is primarily influenced by the methods used to
lift and move these loads. The method of handling fuel will not be
changed since the same equipment and procedures will be used.
Shipping cask movements in the SFP [spent fuel pool] will not be
performed during installation of the new racks. There is no change
to the methods or equipment to be used in moving fuel casks.
Expanding the spent fuel storage capacity does not have a
significant impact on the frequency of occurrence for any accident
previously evaluated.
Therefore, this change will not significantly increase the
probability of occurrence of any accident previously analyzed.
The consequences of a dropped spent fuel assembly in the SFP
have been re-evaluated for the proposed change by analyzing a
potential impact onto the new racks. The results show that the
postulated accident of a fuel assembly striking the new storage
racks will not distort the racks sufficiently to impair their
functionality. The minimum subcriticality margin required by the
current TS (i.e., neutron multiplication factor [keff] less than or
equal to 0.95) will be maintained. The structural damage to the
Reactor Building, pool liner, and fuel assembly resulting from a
dropped fuel assembly striking the pool floor or another assembly
located in the racks is primarily dependent on the mass of the
falling object and the drop height. Since these two parameters are
not changed by the proposed modification, the postulated structural
damage to these items remains unchanged. The radiological dose at
the exclusion area boundary will not be increased since no changes
are being made to in-core hold time or burnup as a result of the
proposed amendment.
Loss of SFP cooling was evaluated. The concern with this event
is a reduction of spent fuel pool water inventory as a result of
boiling in the fuel pool, with the inventory reduction resulting in
an unacceptable increase in dose rates. Loss of spent fuel pool
cooling at CNS is mitigated procedurally by supplying makeup water
to the pool prior to the time that the temperature of the pool
reaches boiling. The thermal-hydraulic analysis for the proposed
license amendment determined, for a complete loss of forced cooling
and a full core discharge, that the minimum time to boil is 4.19
hours. This has been determined to be sufficient time for the
operators to provide alternate means of makeup water to the SFP
before the water begins to boil. Based on this the consequences of a
loss of SFP cooling are not significantly increased.
The consequences of a design basis seismic event are evaluated
on the basis of subsequent fuel damage or compromise of the fuel
storage or building configurations leading to radiological or
criticality concerns. The new racks have been analyzed in their new
configuration and were found to be safe during seismic motion. Fuel
has been determined to remain intact and the storage racks maintain
the fuel and fixed poison configurations subsequent to a seismic
event. The structural capability of the pool and liner will not be
exceeded under the anticipated combinations of dead weight, thermal,
and seismic loads. The Reactor Building structure will remain intact
during a seismic event and will continue to adequately support and
protect the fuel racks, storage array, and pool moderator/coolant.
Therefore, the consequences of a design basis seismic event are not
increased.
The consequence of a fuel misloading accident has been analyzed
for the worst possible storage configuration subsequent to the
proposed modification. It has been determined that the consequences
remain acceptable with respect to the same criteria used previously.
Therefore, the proposed change does not result in a significant
increase in the consequences of a previously evaluated accident.
In summary, 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.
A drop of a fuel assembly onto fuel assemblies stored in the SFP
has been previously analyzed for CNS and is not a new or different
kind of accident. The only event which would represent a new or
different kind of accident is an accidental drop of a rack during
movement in the pool.
Dropping a rack onto stored spent fuel or the pool floor liner,
commonly referred to as a ``heavy load drop,'' is not postulated due
to the defense-in-depth approach to be taken. A lifting rig designed
to meet the requirements of NUREG 0612 [Nuclear Regulatory
Commission technical report designation 0612] and ANSI N 14.6
[American National Standards Institute N 14.6] will be used to
install the new racks. Dropping a new rack onto fuel is precluded by
not allowing the new racks being placed into the SFP to travel over
racks containing fuel assemblies. A rack drop to the pool liner is
not postulated since the lifting components either provide
redundancy in supporting the racks or are designed with safety
margins greater than a factor of ten. Movements of heavy loads over
the pool will comply with the applicable administrative controls and
guidelines (i.e. plant procedures, NUREG 0612, etc.). Therefore, the
rack drop does not represent a new or different kind of accident.
The proposed change does not alter the operation of the plant or
equipment credited for the mitigation of the design basis accidents.
The proposed change does not affect the important parameters
required to ensure safe fuel storage.
In summary, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The function of the spent fuel pool is to store the fuel
assemblies in a subcritical and coolable configuration under
postulated environmental and abnormal loadings, such as an
earthquake or fuel assembly drop. The new rack design meets the
applicable requirements for safe storage and is functionally
compatible with the SFP.
The Holtec Licensing Report was prepared using the guidance of
the applicable provisions of the NRC Guidance entitled, ``OT
Position for Review and Acceptance of Spent Fuel Storage and
Handling Applications.'' The rack materials used are compatible with
the spent fuel assemblies and the SFP environment. The design of the
new racks preserves the proper margin of safety during abnormal
loads, e.g., loads from a seismic event, a dropped assembly, and
tensile loads from a stuck fuel assembly. It has been shown that
such loads will not invalidate the mechanical design and material
selection to safely store fuel in a coolable and subcritical
configuration.
The methodology used in the criticality analysis of the expanded
spent fuel pool
[[Page 70562]]
complies with the appropriate NRC guidelines and the ANSI standards
(Draft GDC 66 [General Design Criterion 66], NUREG 0800, Section
9.1.2, the OT Position for Review and Acceptance of Spent Fuel
Storage and Handling Applications, Reg. Guide 1.13, and ANSI ANS
8.17 [American Nuclear Society 8.17]).
The subcriticality margin (keff) for spent fuel
stored in the SFP is required to be less than or equal to 0.95 under
normal storage, fuel handling, and accident conditions, including
uncertainties. This margin will be maintained with the proposed
increased capacity.
The thermal-hydraulic and cooling evaluation of the pool
determined that the pool can be maintained below the specified
thermal limits under the conditions of the maximum heat load. The
pool temperature will not exceed the design temperature of 150[deg]F
during operation of the cooling systems. The maximum local water
temperature in the hot channel will remain below the boiling point.
The maximum cladding temperature after a loss of cooling remains
less than the current licensing basis value of 350 [deg]F with bulk
boiling in the pool. The stored fuel will not undergo any
significant heat up with blockage of a dropped fuel assembly lying
horizontally on top of the racks. The thermal limits specified for
the evaluations performed to support the proposed change are the
same as those which were used in the previous evaluations.
The time to boiling, in the event of a complete loss of SFP
cooling with a full core discharge, has been reduced from 5 hours to
4.19 hours. However, this has been determined to be sufficient time
for providing makeup to the SFP.
Based on the above it is concluded that the proposed change does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John 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. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: October 19, 2006.
Description of amendment request: The proposed amendment would
revise the surveillance requirements in Technical Specification (TS)
4.1.1, ``Control Rod System,'' to modify the conditions under which
scram time testing (STT) of control rods is required, and add a
requirement to perform STT on a defined portion of control rods, at a
specified frequency, during the operating cycle. The requirement to
test ``eight selected [control] rods'' after a reactor scram or other
outage would be replaced by a requirement to periodically test at least
20 control rods, on a rotating basis, every 180 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change adds new surveillance requirements (SR) to
the MCPR [minimum critical power ratio] Technical Specification (TS)
which requires determination of the MCPR operating limit following
the completion of scram time testing (STT) of the control rods. Use
of the scram speed in determining the MCPR operating limit (i.e.,
Option B) is an alternative to the current method for determining
the operating limit (i.e., Option A). The probability of an accident
previously evaluated is unrelated to the MCPR operating limit that
is provided to ensure no fuel damage results during anticipated
operational occurrences. This is an operational limit to ensure
conditions following an assumed accident do not result in fuel
failure and therefore do not contribute to the occurrence of an
accident.
The proposed change revises allowable conditions for the STT of
non-maintenance affected control rods and eliminates the requirement
to test ``eight [selected] rods'' after a reactor scram or other
outage. The requirement to test ``eight selected rods'' is replaced
by a new SR to perform periodic STT. No active or passive failure
mechanisms that could lead to an accident are affected by this
proposed change and the STT acceptance criteria are not being
revised. Therefore, the proposed change in STT requirements does not
significantly increase the probability or consequences of an
accident previously evaluated.
The proposed change ensures that the appropriate MCPR operating
limit is in place. By implementing the correct MCPR operating limit,
the MCPR SL [safety limit] will continue to be ensured. Ensuring the
MCPR SL is not exceeded will result in prevention of fuel failure.
Therefore, since there is no increase in the potential for fuel
failure, there is no increase in the consequences of any accidents
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 adds a new SR to the MCPR TS which requires
determination of the MCPR operating limit following the completion
of the [STT] of the control rods. The proposed change revises
allowable conditions for the STT of non-maintenance affected control
rods and eliminates the requirement to test ``eight [selected]
rods'' after a reactor scram or other outage. The requirement to
test ``eight selected rods'' is replaced by a new SR to perform
periodic STT. The proposed change does not involve the use or
installation of new equipment. Installed equipment is not operated
in a new or different manner. No new or different system
interactions are created, and no new processes are introduced. No
new failures have been created by the addition of the proposed SR
and the use of the alternate method for determining the MCPR
operating limit. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
Use of Option B for determining the MCPR operating limit will
result in a reduced operating limit in comparison to the use of
Option A. However, a reduction in the operating limit margin does
not result in a reduction in the safety margin. The MCPR SL remains
the same regardless of the method used for determining the operating
limit. The proposed change revises allowable conditions for the STT
of non-maintenance affected control rods and eliminates the
requirement to test ``eight [selected] rods'' after a reactor scram
or other outage. The requirement to test ``eight selected rods'' is
replaced by a new SR to perform periodic STT. No active or passive
failure mechanisms that could adversely impact the consequences of
an accident are affected by this proposed change. All analyzed
transient results remain within the design values for structures,
systems and components. 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: 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: October 23, 2006.
Description of amendment request: The proposed changes to the
technical specifications (TSs) would eliminate the use of the defined
term CORE ALTERATIONS in the TSs.
Basis for proposed no significant hazards consideration
determination:
[[Page 70563]]
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 facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change eliminates the use of the defined term CORE
ALTERATIONS from the Technical Specifications. CORE ALTERATIONS are
not an initiator of any accident previously evaluated except a fuel
handling accident. The revised Technical Specifications that protect
the initial conditions of a fuel handling accident also require the
suspension of movement of irradiated fuel assemblies, which protects
the initial condition of a fuel handling accident.
Therefore, suspension of CORE ALTERATIONS do not affect the
initiators of the accidents previously evaluated and suspension of
CORE ALTERATIONS does not affect the mitigation of the accidents
previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a physical modification of the
plant (i.e., no new or different type of equipment will be
installed) or a significant change in the methods governing normal
plant operation. In addition, the changes do not impose any new or
different requirements. The changes do not alter assumptions made in
the safety analysis. The proposed changes are consistent with the
safety analysis assumptions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
Only two accidents are postulated to occur during plant
conditions where CORE ALTERATIONS may be made: A fuel handling
accident and a boron dilution accident. Suspending movement of
irradiated fuel assemblies prevents a fuel handling accident. Also,
requiring the suspension of CORE ALTERATIONS is redundant to
suspending movement of irradiated fuel assemblies and does not
increase the margin of safety. CORE ALTERATIONS have no effect on a
boron dilution accident. Core components are not involved in the
initiation or mitigation of a boron dilution accident. Therefore,
CORE ALTERATIONS have no effect on the margin of safety related to a
boron dilution accident.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for Licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Acting Branch Chief: L. Raghavan.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: August 4, 2006.
Description of amendment request: The amendments would allow the
use of blind flanges for containment isolation in the containment purge
system supply and exhaust lines, and make corresponding changes to the
Technical Specifications (TSs). The amendments would also consolidate
the containment isolation requirements by moving the requirements of TS
3/4 6.1.7, ``Containment Ventilation System,'' to TS 3/4 6.3.1 (TS 3/4
6.3 for Unit No. 2), ``Containment Isolation Valves.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the Containment purge supply and exhaust
penetrations presents no change in the probability or the
consequence of an accident, since the penetrations continue to
conform to the TS requirements for containment integrity, and will
be appropriately tested as required by 10 CFR 50 Appendix J. The
blind flanges are passive devices not susceptible to an active
failure or malfunction that could result in a loss of isolation or
leakage that exceeds limits assumed in the safety analysis. The
blind flanges are leak rate tested in accordance with the
containment leakage rate testing program. Containment integrity is
not lessened by this change.
The change to the Containment Purge System does not affect the
design basis limit for any fission product barrier.
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 to the Containment purge supply and exhaust
penetrations does not change the function of the system and does not
alter containment integrity. The penetrations continue to conform to
the TS requirements for containment integrity and will be
appropriately tested as required by 10 CFR 50 Appendix J. No new
accident scenarios, failure mechanisms, or limiting single failures
are introduced as a result of the proposed changes.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
The proposed change will not alter any assumptions, initial
conditions or results specified in any accident analysis. The
Containment purge supply and exhaust penetrations will continue to
conform to the TS requirements for containment integrity, and will
be appropriately tested as required by 10 CFR 50 Appendix J. The
blind flanges are passive devices not susceptible to an active
failure or malfunction that could result in a loss of isolation or
leakage that exceeds limits assumed in the safety analysis. The
blind flanges are leak rate tested in accordance with the
containment leakage rate testing program. Containment integrity is
not lessened by this change. Therefore, there is no 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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Branch Chief: Harold K. Chernoff.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: October 3, 2006.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) and licensing basis to support
the resolution of the Nuclear Regulatory Commission's (NRC's) Generic
Safety Issue (GSI) 191, assessment of debris accumulation on
containment sump performance and its impact on emergency recirculation
during an accident, and NRC Generic Letter (GL) 2004-02.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
[[Page 70564]]
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response: No.
The proposed changes include a physical alteration to the RS
system to start the inside and outside [Recirculation Spray] RS
pumps on [Refueling Water Storage Tank] RWST Level Low coincident
with High High containment pressure. The RS system is used for
accident mitigation only, and changes in the operation of the RS
system cannot have an impact on the probability of an accident. The
other changes do not affect equipment and are not accident
initiators. The RWST Level Low instrumentation will comply with all
applicable regulatory requirements and design criteria (e.g., train
separation, redundancy, and single failure). Therefore, the design
functions performed by the RS system are not changed.
Delaying the start of the RS pumps creates more challenging
long-term containment pressure and temperature profiles. The
environmental qualification of safety-related equipment inside
containment was confirmed to be acceptable, and accident mitigation
systems will continue to operate within design temperatures and
pressures. Delaying the RS pump start reduces the emergency diesel
generator loading early during a design basis accident, and
staggering the RS pump start avoids overloading on each emergency
bus. The reduction in iodine removal efficiency during the delay
period is offset by changes to other assumptions in the [loss-of-
coolant accident] LOCA dose analysis. The predicted offsite doses
and control room doses following a design basis LOCA remain within
regulatory limits.
The [Updated Final Safety Analysis Report] UFSAR safety analysis
acceptance criteria continue to be met for the proposed changes to
the RS pump start method, the proposed TS containment air partial
pressure limits, the proposed TS containment temperature limit, the
implementation of the GOTHIC containment analysis methodology, the
proposed change to the [safety injection] SI [recirculation mode
transfer] RMT allowable values, and the changes to the LOCA dose
consequences analyses. Based on this discussion, the proposed
amendments do not increase the probability or consequence of an
accident previously evaluated.
2. Does the proposed license amendment create the possibility of
a new or different kind of accident from any accident previously
identified?
Response: No.
The proposed change alters the RS pump circuitry by initiating
the start sequence with a new RWST Level Low signal instead of a
timer after the High High containment pressure setpoint is reached.
The timers for the inside RS pumps will be used to sequence pump
starts and preclude diesel generator overloading. The RS pump
function is not changed. The RWST Level Low instrumentation will be
included as part of the Engineered Safety Features Actuation System
(ESFAS) instrumentation in the North Anna TS and will be subject to
the ESFAS surveillance requirements. The design of the RWST Level
Low instrumentation complies with all applicable regulatory
requirements and design criteria. The failure modes have been
analyzed to ensure that the RWST Level Low circuitry can withstand a
single active failure without affecting the RS system design
functions. The RS system is an accident mitigation system only, so
no new accident initiators are created.
The remaining changes to the containment analysis methodology,
the containment air partial pressures, the maximum containment
temperature operating limit, the TS allowable values for SI RMT, and
the LOCA [alternate source term] AST analysis basis do not impact
plant equipment design or function. Together, the changes assure
that there is adequate margin available to meet the safety analysis
criteria and that dose consequences are within regulatory limits.
The proposed changes do not introduce failure modes, accident
initiators, or malfunctions that would cause a new or different kind
of accident. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously identified.
3. Does the proposed license amendment involve a significant
reduction in a margin of safety?
Response: No.
The changes to the actuation of the RS pumps and the increased
containment air partial pressure have created an adverse effect on
the containment response analyses and the LOCA dose analysis.
Analyses have been performed that show the containment design basis
limits are satisfied and the post-LOCA offsite and control room
doses meet the required criteria for the proposed changes to the
containment analysis methodology, the RS pump start method, the TS
containment air partial pressure limits, the TS containment
temperature maximum limit, the TS allowable values for SI RMT, and
the LOCA AST bases. Therefore, the proposed amendment does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Branch Chief: Evangelos C. Marinos.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: November 16, 2006.
Description of amendment request: The proposed amendments would add
a reference in Technical Specification (TS) 6.2.C, ``Core Operating
Limits Report (COLR),'' to permit the use of the Westinghouse Best-
Estimate Large Break Loss of Coolant Accident (BE-LBLOCA) analysis
methodology using the Automated Statistical Treatment of Uncertainty
Method (ASTRUM) for the analysis of LBLOCA.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The probability of occurrence or the consequences of an
accident previously evaluated are not significantly increased.No
physical plant changes are being made as a result of using the
Westinghouse Best Estimate Large Break LOCA (BE-LBLOCA) analysis
methodology. The proposed TS change simply involves updating the
references in TS 6.2.C, Core Operating Limits Report (COLR), to
reference the Westinghouse BE-LBLOCA analysis methodology. The
consequences of a LOCA are not being increased, since the analysis
has shown that the Emergency Core Cooling System (ECCS) is designed
such that its calculated cooling performance conforms to the
criteria contained in 10 CFR 50.46, ``Acceptance criteria for
emergency core cooling systems for light-water nuclear power
reactors.'' No other accident consequence is potentially affected by
this change.
All systems will continue to be operated in accordance with
current design requirements under the new analysis, therefore no new
components or system interactions have been identified that could
lead to an increase in the probability of any accident previously
evaluated in the Updated Final Safety Analysis Report (UFSAR). No
changes were required to the Reactor Protection System (RPS) or
Engineering Safety Features (ESF) setpoints because of the new
analysis methodology.
An analysis of the LBLOCA accident for Surry Units 1 and 2 has
been performed with the Westinghouse BE-LBLOCA analysis methodology
using ASTRUM. The analysis was performed in compliance with all the
NRC conditions and limitations as identified in WCAP-16009-P-A.
Based on the analysis results, it is concluded that the Surry Units
1 and 2 continue to maintain a margin of safety to the limits
prescribed by 10 CFR 50.46.
There are no changes to assumptions of the radiological dose
calculations. Hence, there is no increase in the predicted
radiological consequences of accidents postulated in the UFSAR.
Therefore, neither the probability of occurrence nor the
consequences of an accident previously evaluated is significantly
increased.
2. The possibility for a new or different type of accident from
any accident previously evaluated is not created.
The use of the Westinghouse BE-LBLOCA analysis methodology with
ASTRUM does
[[Page 70565]]
not impact any of the applicable design criteria and all pertinent
licensing basis criteria will continue to be met. Demonstrated
adherence to the criteria in 10 CFR 50.46 precludes new challenges
to components and systems that could introduce a new type of
accident. Safety analysis evaluations have demonstrated that the use
of Westinghouse BE-LBLOCA analysis methodology with ASTRUM is
acceptable. All design and performance criteria will continue to be
met and no new single failure mechanisms will be created. The use of
the Westinghouse BE-LBLOCA analysis methodology with ASTRUM does not
involve any alteration to plant equipment or procedures that would
introduce any new or unique operational modes or accident
precursors. Furthermore, no changes have been made to any RPS or ESF
actuation setpoints. Based on this review, it is concluded that no
new accident scenarios, failure mechanisms, or limiting single
failures are introduced as a result of the proposed changes.
Therefore, the possibility for a new or different kind of
accident from any accident previously evaluated is not created.
3. The margin of safety is not significantly reduced.
It has been shown that the analytical technique used in the
Westinghouse BE-LBLOCA analysis methodology using ASTRUM
realistically describes the expected behavior of the reactor system
during a postulated LOCA. Uncertainties have been accounted for as
required by 10 CFR 50.46. A sufficient number of LOCAs with
different break sizes, different locations, and other variations in
properties have been considered to provide assurance that the most
severe postulated LOCAs have been evaluated. The analysis has
demonstrated that all acceptance criteria contained in 10 CFR 50.46
continue to be satisfied.
Therefore, it is concluded that this change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385
NRC Branch Chief: Evangelos C. Marinos
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of application for amendment: December 1, 2005.
Brief description of amendment: The amendment revised Technical
Specification 3.6.4.1, ``Secondary Containment.'' Specifically, the
amendment revised Surveillance Requirement (SR) 3.6.4.1.4 and SR
3.6.4.1.5 to clarify their intent with respect to secondary containment
boundary integrity.
Date of issuance: November 17, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 175.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specification Surveillance Requirements and License.
Date of initial notice in Federal Register: March 28, 2006 (71 FR
15481).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 17, 2006.
No significant hazards consideration comments received: No.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: September 26, 2006, as
supplemented by the letter dated November 3, 2006.
Brief description of amendments: The amendments revise TS 3.7.2,
``Main Steam Isolation Valves (MSIVs),'' to include specific
requirements for the MSIV actuator trains.
Date of issuance: November 17, 2006.
Effective date: Effective as of the date of issuance to be
implemented within 10 days from the date of issuance.
Amendment Nos.: Unit 1--163, Unit 2--163, Unit 3--163.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 5, 2006 (71 FR
58879). The supplemental letter dated November 3, 2006, provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 17, 2006.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: February 27, 2006.
Brief description of amendments: The amendments revise Technical
[[Page 70566]]
Specification 4.2.1, ``Fuel Assemblies,'' to permit up to four lead
fuel assemblies (LFAs) with advanced cladding material to be re-
inserted into either the Unit 1 or Unit 2 core for the next operating
cycle, which is Cycle 19 for Unit 1 and Cycle 17 for Unit 2. Two of
these LFAs were manufactured by Westinghouse Electric Company and
contain a limited number of fuel rods with advanced zirconium-based
alloys. The other two LFAs were manufactured by Framatome ANP, Inc.
with fuel rod cladding material classified as M5TM alloy.
These LFAs were originally inserted into the Unit 2 core in April 2003
(Operating Cycles 15 and 16) and are scheduled to be discharged during
the 2007 refueling outage.
Date of issuance: November 16, 2006.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 280 and 257.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the License and Technical Specifications.
Date of initial notice in Federal Register: March 28, 2006 (71 FR
15482).
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated November 16, 2006.
No significant hazards consideration comments received: No
Exelon Generation Company, LLC, Docket No. 50-249, Dresden Nuclear
Power Station, Unit 3, Grundy County, Illinois
Date of application for amendment: July 21, 2006, as supplemented
by letter dated October 19, 2006.
Brief description of amendment: The amendment revised the values of
the safety limit minimum critical power ratio in Technical
Specification Section 2.1.1, ``Reactor Core SLs [Safety Limits].''
Date of issuance: November 7, 2006.
Effective date: As of the date of issuance and shall be implemented
prior to startup for cycle 20.
Amendment Nos.: 213.
Renewed Facility Operating License Nos. DPR-19 and DPR-25: The
amendment revised the Technical Specifications and License.
Date of initial notice in Federal Register: August 29, 2006 (71 FR
51228). The October 19, 2006 supplement provided additional clarifying
information that did not expand the scope of the application as
originally noticed, and did not change the NRC staff's original
proposed no significant hazards consideration determination published
in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 7, 2006.
No significant hazards consideration comments received: No.
Omaha Public Power District (OPPD), Docket No. 50-285, Fort Calhoun
Station, Unit No. 1 (FCS), Washington County, Nebraska
Date of amendment request: August 21, 2006, as supplemented on
September 6 and October 10, 2006.
Brief description of amendment: The amendment changed the Technical
Specifications (TSs) to: (1) Revise TS Section 2.3(4) to change the
reactor containment building sump buffering agent from trisodium
phosphate to sodium tetraborate and change the TS section title to
``Containment Sump Buffering Agent Specification and Volume
Requirement,'' (2) revise TS 3.6(2)d to require a volume of sodium
tetraborate that is within an area of acceptable operation, as shown in
TS Figure 2-3, and (3) an administrative correction to TS 3.6(2)d(i).
The amendment allows OPPD to replace the trisodium phosphate in the
containment with sodium tetraborate. Changes were also made to the
corresponding TS Bases. The TS changes are approved for Cycle 24 only,
ending in the spring 2008 refueling outage.
Date of issuance: November 13, 2006.
Effective date: As of its date of issuance and shall be implemented
within 90 days of issuance.
Amendment No.: 247.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: August 30, 2006 (71 FR
51646). The September 6 and October 10, 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.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated November 13, 2006.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: November 3, 2005, as
supplemented by letters dated May 1, August 15, and October 5, 2006.
Brief description of amendments: The amendments revised Technical
Specification Section 5.5.2.11 to modify the definitions of steam
generator tube ``Repair Limit'' and ``Tube Inspection.'' The changes
define the extent of the required tube inspections and repair criteria
within the tubesheet regions.
Date of issuance: November 9, 2006.
Effective date: As of its date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 2--206; Unit 3--198.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 6, 2005 (70 FR
72676). The May 1, August 15, and October 5, 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 no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 9, 2006.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: July 14, 2006.
Brief description of amendments: The amendments deleted duplicative
notifications, reporting, and restart requirements if a safety limit
was violated; replaced plant-specific position titles with generic
position titles; and additional administrative changes.
Date of issuance: November 15, 2006.
Effective date: As of date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 2--207; Unit 3--199.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
[[Page 70567]]
Date of initial notice in Federal Register: September 12, 2006 (71
FR 53720).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 15, 2006.
No significant hazards consideration comments received: No.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: May 11, 2006.
Brief description of amendment: The amendment revised Surveillance
Requirements (SRs) 3.7.2.1, 3.7.3.1, and 3.7.3.3 on verifying the
closure time of the main steam isolation valves (MSIVs), main feedwater
regulating valves (MFRVs), main feedwater regulating valve bypass
valves (MFRVBVs), and main feedwater isolation valve (MFIVs) in the
Technical Specifications (TS). These valves are the Main Steam and Main
Feedwater System isolation valves. The revisions replace (1) the
specified maximum acceptable valve closure time for the MSIVs, MFRVs,
and MFRVBVs, and (2) TS Figure 3.7.3-1, which shows acceptable valve
closure times for the MFIVs, by the reference to the valve closure time
is verified to be ``within limits.'' The maximum acceptable valve
closure times for the MFRVs and MFRVBVs, and TS Figure 3.7.3-1 are now
located in the TS Bases. The maximum acceptable valve closure time for
the MSIV is already in the TS Bases.
Date of issuance: November 15, 2006.
Effective date: Effective as of its date of issuance, and shall be
implemented within 90 days of the date of issuance.
Amendment No.: 176.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 20, 2006 (71 FR
35461).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 15, 2006.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
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
[[Page 70568]]
CFR Part 2. Interested persons should consult a current copy of 10 CFR
2.309, which is available at the Commission's PDR, located at One White
Flint North, Public File Area 01F21, 11555 Rockville Pike (first
floor), Rockville, Maryland, and electronically on the Internet at the
NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If
there are problems in accessing the document, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected].
If a request for a hearing or petition for leave to intervene is filed
by the above date, the Commission or a presiding officer designated by
the Commission or by the Chief Administrative Judge of the Atomic
Safety and Licensing Board Panel, will rule on the request and/or
petition; and the Secretary or the Chief Administrative Judge of the
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
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\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly avaialble 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.
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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).
Pacific Gas and Electric Company, Docket No. 50-275, Diablo Canyon
Nuclear Power Plant, Unit No. 1, San Luis Obispo County, California
Date of application for amendment: October 18, 2006, as
supplemented on November 2, 2006.
Brief description of amendment: The amendment revised Technical
Specification (TS) Section 3.8.4, ``DC Sources--Operating,'' Condition
B to extend the completion time (CT) to restore an inoperable vital
battery from 2 hours to 4 hours for the current operating Cycle 14,
provided certain required actions are taken. The extended CT would
allow sufficient time to correct a degraded condition on the station
Vital Battery 1-1.
Date of issuance: November 15, 2006
Effective date: As of its date of issuance and shall be implemented
within 7 days of the date of issuance.
Amendment No.: 190
Facility Operating License No. DPR-80: The amendment revised the
Technical Specifications and license.
Public comments requested as to proposed no significant hazards
[[Page 70569]]
consideration (NSHC): Yes. An individual 14-day Notice of Consideration
of Issuance of Amendment to Facility Operating License was published on
October 27, 2006 (71 FR 63040) in the Federal Register. 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 December 26, 2006, but
indicated that if the Commission makes a final NSHC determination, any
such hearing would take place after issuance of the amendment.
The November 2, 2006, supplemental letter provided additional
information that clarified the application, and did not expand the
scope of the application as originally noticed.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated November 15, 2006.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Branch Chief: David Terao
Dated at Rockville, Maryland, this 22nd day of November 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E6-20329 Filed 12-4-06; 8:45 am]
BILLING CODE 7590-01-P