[Federal Register Volume 72, Number 1 (Wednesday, January 3, 2007)]
[Notices]
[Pages 147-159]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-22492]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a. (2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from December 8, 2006 to December 21, 2006. The
last biweekly notice was published on December 19, 2006 (71 FR 75987).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazard 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 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board 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.
[[Page 148]]
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: September 28, 2006.
Description of amendment request: The proposed amendment would
modify technical specification (TS) requirements of TS 3.8.3, ``Diesel
Fuel Oil,'' to include a new Condition A with associated Required
Action and Completion Time. The proposed Condition A allows the main
fuel oil storage tank to be inoperable for up to 14 days for the
purpose of performing inspection, cleaning, or repair activities.
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 does not alter the assumption of the
accident analyses or the Technical Specification Bases. The
inclusion of provisions to permit internal inspection of the main
fuel oil storage tank during plant operation does not impact the
availability of the EDGs to perform their intended safety function.
Furthermore, while the main fuel oil storage tank is out of service,
the availability of on-site and off-site fuel oil sources ensures
that an adequate supply of fuel oil remains available. Therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical change to the
design of the Diesel Fuel Oil System, nor does it alter the
assumptions of the accident analyses. The inclusion of provisions to
permit internal inspection and cleaning of the main fuel oil storage
tank during plant operation does not introduce any new failure
modes. 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?
[[Page 149]]
Response: No.
The proposed change alters the method of operation of the Diesel
Fuel Oil System. However, the availability of the EDGs to perform
their intended safety function is not impacted and the assumptions
of the accident analyses are not altered. Therefore, the proposed
change does not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Acting Branch Chief: D. Pickett.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: November 27, 2006.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TSs) by relocating references to
specific American Society for Testing and Materials (ASTM) standards
for fuel oil testing to licensee-controlled documents and adding
alternate criteria to the ``clear and bright'' acceptance test for new
fuel oil.
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 by a reference to a generic analysis published in the
Federal Register on February 22, 2006 (71 FR 9179), which is presented
below:
1. Does the proposed change involve a significant increase in
the probability or consequences of any accident previously
evaluated?
Response: No.
The proposed changes relocate the specific ASTM standard
references from the Administrative Controls Section of TS to a
licensee-controlled document. Requirements to perform testing in
accordance with applicable ASTM standards are retained in the TS as
are requirements to perform surveillances of both new and stored
diesel fuel oil. Future changes to the licensee-controlled document
will be evaluated pursuant to the requirements of 10 CFR 50.59,
``Changes, tests and experiments,'' to ensure that such changes do
not result in more than a minimal increase in the probability or
consequences of an accident previously evaluated. In addition, the
``clear and bright'' test used to establish the acceptability of new
fuel oil for use prior to addition to storage tanks has been
expanded to recognize more rigorous testing of water and sediment
content. Relocating the specific ASTM standard references from the
TS to a licensee-controlled document and allowing a water and
sediment content test to be performed to establish the acceptability
of new fuel oil will not affect nor degrade the ability of the
emergency diesel generators (DGs) to perform their specified safety
function. Fuel oil quality will continue to meet ASTM requirements.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, and
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not adversely
affect the ability of structures, systems, and components (SSCs) to
perform their intended safety function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of any accident previously evaluated.
Further, the proposed changes do not increase the types and amounts
of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational/public
radiation exposures.
Therefore, the changes do not involve a significant increase in
the probability or consequences of any accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes relocate the specific ASTM standard
references from the Administrative Controls Section of TS to a
licensee-controlled document. In addition, the ``clear and bright''
test used to establish the acceptability of new fuel oil for use
prior to addition to storage tanks has been expanded to allow a
water and sediment content test to be performed to establish the
acceptability of new fuel oil. 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. The requirements retained in the TS continue to require
testing of the diesel fuel oil to ensure the proper functioning of
the DGs.
Therefore, the changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes relocate the specific ASTM standard
references from the Administrative Controls Section of TS to a
licensee-controlled document. Instituting the proposed changes will
continue to ensure the use of applicable ASTM standards to evaluate
the quality of both new and stored fuel oil designated for use in
the emergency DGs. Changes to the licensee-controlled document are
performed in accordance with the provisions of 10 CFR 50.59. This
approach provides an effective level of regulatory control and
ensures that diesel fuel oil testing is conducted such that there is
no significant reduction in a margin of safety.
The ``clear and bright'' test used to establish the
acceptability of new fuel oil for use prior to addition to storage
tanks has been expanded to allow a water and sediment content test
to be performed to establish the acceptability of new fuel oil. The
margin of safety provided by the DGs is unaffected by the proposed
changes since there continue to be TS requirements to ensure fuel
oil is of the appropriate quality for emergency DG use. The proposed
changes provide the flexibility needed to improve fuel oil sampling
and analysis methodologies while maintaining sufficient controls to
preserve the current margins of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: David G. Pettinari, Legal Department, 688
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
NRC Branch Chief: L. Raghavan.
Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: April 11, 2006.
Description of amendment request: The proposed amendments would
(Item 1) revise the Technical Specifications (TSs) and delete the
license conditions related to steam generator (SG) tube integrity and
(Item 2) revise an organizational description in TS 5.2.1 that is
solely administrative in nature and unrelated to the SG tube integrity
TSs.
The changes related to SG tube integrity 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:
[[Page 150]]
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
(Item 1) SG Tube Integrity
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.
(Item 2) Organization Description Revision in TS 5.2.1
The proposed change revises an organizational description in TS
5.2.1 to conform to an application for consent to the indirect transfer
of control of the renewed facility operating licenses. The proposed
change does not affect the operation of any equipment, and is solely
administrative in nature; therefore, the proposed change has no impact
on any accident probabilities or consequences.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
(Item 1) SG Tube Integrity
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.
(Item 2) Organization Description Revision in TS 5.2.1
There are no new accident causal mechanisms created as a result of
this proposed change. No changes are being made to the plant that will
introduce any new accident causal mechanisms. This change is solely
administrative in nature and does not impact any plant systems that are
accident initiators; therefore, no new accident types are being
created.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
(Item 1) SG Tube Integrity
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
[[Page 151]]
and are an improvement over the requirements in the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the TS.
(Item 2) Organization Description Revision in TS 5.2.1
Margin of safety is related to confidence in the ability of the
fission product barriers to perform their design functions during and
following an accident situation. This proposed change is solely
administrative in nature and does not affect the performance of the
barriers. Consequently, no safety margins will be impacted.
Attorney for licensee: Lisa F. Vaughn, Associate General Counsel
and Managing Attorney, Duke Energy Carolinas, LLC, 526 South Church
Street EC07H, Charlotte, NC 28202.
NRC Branch Chief: Evangelos C. Marinos.
Duke Power Company LLC, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: November 16, 2006.
Description of amendment request: The proposed amendments would
authorize revision to revise the Updated Final Safety Analysis Report
(UFSAR) to describe the flood protection measures for the auxiliary
building.
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. This License Amendment Request (LAR) proposes the use of a
realistic seismic evaluation of the Auxiliary Building sprinkler
system (high pressure service water) piping which demonstrates that
these non-Category I (non-seismic) self-actuating sprinkler systems
will not fail during a Maximum Hypothetical Earthquake (MHE) and
clarifies Duke's commitment toward Auxiliary Building flood
protection measures in the Updated Final Safety Analysis Report
(UFSAR). The proposed change does not affect any Chapter 15 accident
analyses. Operation in accordance with the amendment authorizing
this change would not involve any accident initiation sequences or
change the consequences of any accident analyzed.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. This LAR proposes the use of a realistic seismic evaluation
of the Auxiliary Building sprinkler system (high pressure service
water) piping which demonstrate that these non-Category I (non-
seismic) self-actuating sprinkler systems will not fail during a MHE
and clarifies Duke's commitment toward Auxiliary Building flood
protection measures in the UFSAR. Operation in accordance with this
proposed amendment will not result in a change in the parameters
governing plant operation and will not generate any new accident
initiators. 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?
No. This LAR proposes the use of a realistic seismic evaluation
of the Auxiliary Building sprinkler system (high pressure service
water) piping, which demonstrates that these non-Category I (non-
seismic) self-actuating sprinkler systems will not fail during a MHE
and clarifies Duke's commitment toward Auxiliary Building flood
protection measures in the UFSAR. Operation in accordance with this
proposed amendment will not result in a change in the parameters
governing plant operation and will not affect any Chapter 15
accident analyses. 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: Ms. Lisa F. Vaughn, Associate General
Counsel and Managing Attorney, Duke Energy Carolinas, LLC, 526 South
Church Street, Charlotte, NC 28202.
NRC Branch Chief: Evangelos C. Marinos.
Duke Power Company LLC, et al., Docket Nos. 50-269, 50-270, and 50-287,
Oconee Nuclear Station, Units 1, 2, and 3, Oconee 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
[[Page 152]]
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: Lisa F. Vaughn, Associate General Counsel
and Managing Attorney, Duke Energy Carolinas, LLC, 526 South Church
Street, EC07H, Charlotte, NC 28202.
NRC Branch Chief: Evangelos C. Marinos.
Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of amendment request: September 26, 2006.
Description of amendment request: The proposed amendment would
allow up to eight AREVA NP Inc. Modified Advanced Mark-BW(A) fuel
assemblies containing M5 alloy to be placed in nonlimitng Braidwood
Station, Unit No. 1 core regions for evaluation during Cycle 14, 15,
and 16. The proposed amendment would also remove all references to
Joseph Oat spent fuel storage racks that have been physically removed
from the spent fuel pool.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS [technical specification] change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The AREVA Advanced Mark-BW(A) fuel is similar in design to the
Westinghouse fuel that will be co-resident in the core. The Advanced
Mark-BW(A) fuel assemblies are also similar in design to the
Advanced Mark-BW assemblies using M5 alloy material for the
cladding, structural tubing, and grids generically approved for use
in Westinghouse 3- and 4-loop designed pressurized water reactors
with 17 x 17 fuel rod arrays. The AREVA Advanced Mark-BW(A) fuel
assemblies will be placed in nonlimiting regions (i.e., locations)
of the core. The Cycle 14, 15, and 16 reload designs will meet all
applicable design criteria. EGC [Exelon Generation Company, LLC]
will use the NRC-approved standard reload design models and methods
to demonstrate that all applicable design criteria will be met.
Evaluations will be performed as part of the cycle specific reload
safety analysis for the operation of the AREVA Advanced Mark-BW(A)
fuel to confirm that the acceptance criteria of the existing safety
analyses continue to be met. Operation of the AREVA Advanced Mark-
BW(A) fuel will not significantly increase the predicted
radiological consequences of accidents postulated in the Updated
Final Safety Analysis Report.
The proposed change regarding removal of all references in TS to
the Joseph Oat spent fuel racks is administrative and deletes
unnecessary wording relating to equipment that is physically removed
from the Braidwood Station spent fuel pool and therefore does not
alter the design, configuration, operation, or function of any plant
system, structure or component. As a result, the administrative
change does not affect the outcome of any previously evaluated
accidents.
Based on the above discussion, the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The AREVA Advanced Mark-BW(A) fuel is similar in design to the
Westinghouse fuel that will be co-resident in the core. The Advanced
Mark-BW(A) fuel assemblies are also similar in design to the
Advanced Mark-BW assemblies using M5 alloy material for the
cladding, structural tubing, and grids generically approved for use
in Westinghouse 3- and 4-loop designed pressurized water reactors
with 17 x 17 fuel rod arrays. The Braidwood Station, Unit [No.] 1
cores in which the fuel operates will be designed to meet all
applicable design criteria and ensure that all pertinent licensing
basis criteria are met. Demonstrated adherence to these standards
and criteria precludes new challenges to components and systems that
could introduce a new type of accident. The reload core designs for
the cycles in which the AREVA Advanced Mark-BW(A) fuel will operate
will demonstrate that the use of up to eight AREVA Advanced Mark-
BW(A) fuel assemblies in nonlimiting core regions (i.e., locations)
is acceptable. The relevant design and performance criteria will
continue to be met and no new single failure mechanisms will be
created. The use of AREVA Advanced Mark-BW(A) fuel does not involve
any alteration to plant equipment or procedures that would introduce
any new or unique operational modes or accident precursors.
The proposed change regarding removal of all references in TS to
Joseph Oat spent fuel racks is administrative and deletes
unnecessary wording relating to equipment that is physically removed
from the Braidwood Station spent fuel pool and therefore does not
alter the design, configuration, operation, or function of an plant
system, structure or component. As a result, the administrative
change does not create any new or different kind of accident.
Based on this evaluation, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
Operation of Braidwood Station, Unit [No.] 1 with up to eight
AREVA Advanced Mark-BW(A) fuel assemblies in nonlimiting core
regions (i.e., locations) does not change the performance
requirements on any system or component such that any design
criteria will be exceeded. The normal limits on core operation
defined in the Braidwood Station TS will remain applicable for the
use of up to eight AREVA Advanced Mark-BW(A) fuel assemblies during
Cycles 14, 15, and 16. The reload core designs for the cycles in
which the AREVA Advanced Mark-BW(A) fuel will operate will
specifically evaluate any pertinent differences, including both
mechanical design differences and the past irradiation history,
between the AREVA Advanced Mark-BW(A) fuel product, and the
Westinghouse fuel product that will be co-resident in the core. The
use of up to eight AREVA Advanced Mark-BW(A) fuel assemblies will be
specifically evaluated during the reload design process using reload
design models and methods as approved by the NRC.
The proposed change regarding removal of all references in TS to
Joseph Oat spent fuel racks is administrative and deletes
unnecessary wording relating to equipment that is physically removed
from the Braidwood Station spent fuel pool and therefore does not
alter the design, configuration, operation, or function of an plant
system, structure or component. As a result, the administrative
change does not affect the ability of any operable structure,
system, or component to perform its designated safety function.
Based on this evaluation, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Michael L. Marshall, Jr.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: October 19, 2006.
Description of amendment request: The proposed amendments would
[[Page 153]]
revise Technical Specification 4.6.2.1.d to allow the frequency of air
or smoke flow testing of the containment spray nozzles to be reduced
from 10 years to an activity-related frequency following maintenance
that could cause a blockage.
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 change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed change revises the surveillance frequency from
once per 10 years to following activities that could result in
nozzle blockage. The containment spray system nozzles are passive
components and are not considered as an initiator of any analyzed
event. The proposed change will not impact the ability of the
containment spray system to mitigate the consequences of an
accident. Industry experience indicates that containment spray
systems of similar design are highly reliable and not susceptible to
plugging due to the open design of the nozzles, the location of the
nozzles high in the containment dome, and the corrosion resistant
materials used for construction of the system. The alternative
frequency of this surveillance has no impact on the probability of
failure of associated active systems. Therefore, there is no
significant increase in the probability or consequences of
previously evaluated accidents due to the extended surveillance
frequency.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. The proposed amendment provides an alternative frequency for
performance of the spray nozzle surveillance test. The containment
spray nozzles are used for accident mitigation only. Potential
unidentified blockage of the containment spray nozzles will not
result in the initiation of an accident. The change does not involve
a physical alteration of the plant nor does it result in an
operational condition different from that which has already been
considered in the accident analyses. Therefore, the change does not
create the possibility of a new or different kind of accident or
malfunction from any accident previously evaluated.
3. Does this change involve a significant reduction in margin of
safety?
No. The alternative frequency of spray nozzle testing has no
significant impact on the consequences of any analyzed accident and
does not significantly change the failure probability of any
equipment that provides protection for the health and safety of the
public. The containment spray system will continue to be capable of
maintaining containment temperature and pressure below design
values. Therefore, there is no significant reduction in the margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Douglas V. Pickett (Acting).
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: October 19, 2006.
Description of amendment request: The proposed amendments would
revise various Technical Specifications (TSs) to address requirements
that should have been changed as part of previously approved
amendments. These amendments included TS changes regarding relocation
of administrative requirements to licensee controlled programs such as
the Topical Quality Assurance Report (TQAR), handling of recently
irradiated fuel in accordance with TS Task Force change traveler TSTF-
51, and Auxiliary Feedwater Actuation System (AFAS) trip and bypass
requirements. The proposed amendments also correct some typographical
errors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
These proposed license amendments require no plant hardware or
operational modifications. The proposed changes either correct
various administrative errors or incorporate changes that have been
justified by previously approved license amendments and should have
been made as part of those submittals. Therefore, operation of the
facility in accordance with the proposed amendments would 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 modifications to either plant hardware or operational
procedures are required to support these proposed license
amendments; hence, no new failure modes are created. The proposed
changes either correct various administrative errors or incorporate
changes that have been justified by previously approved license
amendments and should have been made as part of those submittals.
Therefore, operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The TS corrections proposed by these license amendments are
administrative in nature in that they either correct typographical
errors (e.g., letter dates and transient limits) or are justified by
previous license amendments (i.e., relocation of administrative
programs to the TQAR, the implementation of TSTF-51 for recently
irradiated fuel, and correct inconsistencies introduced by AFAS trip
and bypass requirements). Therefore, operation of the facility in
accordance with the proposed amendment would 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: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Branch Chief: Douglas V. Pickett (Acting).
Indiana Michigan Power Company (I&M), Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit 1 and 2, Berrien County, Michigan
Date of amendment request: November 3, 2006.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TS) to reflect a proposed plant
modification that will replace the reactor coolant system resistance
temperature detectors (RTDs) and bypass piping with fast response
thermowell detectors mounted directly in the primary loop piping. The
specific TS requirements affected include the notes in Unit 2 TS
surveillance requirement for channel calibration of the overtemperature
differential temperature (OT[Delta]T) and overpower differential
temperature (OP[Delta]T) reactor trip system functions. The proposed
change also affects the Unit 1 and Unit 2 TS allowable values for
OT[Delta]T and OP[Delta]T reactor trip systems functions.
Basis for proposed no significant hazards consideration
determination:
[[Page 154]]
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 of occurrence or consequences of an accident
previously evaluated?
Response: No.
The resistance temperature detectors (RTD) bypass system is the
hardware associated with Reactor Coolant System instrumentation
having control, indication, and protection functions. The RTD bypass
system is not considered a precursor to any previously analyzed
accident. The system is relied upon to mitigate the consequences of
some accidents. The new system replacing the RTD bypass system will
perform the same control, indication, and protection functions, and,
similarly, will not be considered a precursor to any accident. The
capability of the system to mitigate the consequences of the
previously analyzed accidents will not be significantly affected.
Therefore, replacement of the existing RTD bypass system with the
new system will not increase the probability of occurrence of an
accident, and will not increase 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 replacement of the existing RTD bypass with the new system
would not create new failure modes, and the replacement system is
not an initiator of any new or different kind of accident. The
proposed deletion of the note in Technical Specification (TS)
Surveillance Requirement 3.3.1.15, and proposed changes to Allowable
Values in TS Table 3.3.1-1 do not affect the interaction of the
replacement system with any system whose failure or malfunction can
initiate an accident. 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.
Margins of safety are established in the design of components,
the configuration of components to meet certain performance
parameters, and in the models and associated assumptions used to
analyze the system's performance. The replacement system will
continue to perform the same temperature detection function to the
same level of reliability as defined in the Donald C. Cook Nuclear
Plant Updated Final Safety Analysis Report. Therefore, the proposed
amendment does not involve a significant reduction in a margin of
safety.
The Nuclear Regulatory Commission (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 requests involve no
significant hazards consideration.
Attorney for licensee: Kimberly Harshaw, Esquire, One Cook Place,
Bridgman, MI 49106.
NRC Branch Chief: L. Raghavan.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: December 14, 2006.
Description of amendment requests: The proposed amendments would
delete Section 2.G of the Diablo Canyon Power Plant Facility Operating
Licenses, which requires reporting of violations of the requirements in
Sections 2.C, 2.E, and 2.F of the Facility Operating License.
The NRC staff issued a notice of opportunity to comment in the
Federal Register on August 29, 2005 (70 FR 51098), on possible
amendments to eliminate the license condition involving reporting of
violations of other requirements (typically in License Condition 2.C)
in the operating license, 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 model for referencing in license
amendment applications in the Federal Register on November 4, 2005 (70
FR 67202). The licensee affirmed the applicability of the NSHC
determination in its application dated December 14, 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change involves the deletion of a reporting
requirement. The change does not affect plant equipment or operating
practices and therefore does not significantly increase the
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The proposed change is administrative in that it deletes a
reporting requirement. The change does not add new plant equipment,
change existing plant equipment, or affect the operating practices
of the facility. Therefore, the change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change deletes a reporting requirement. The change
does not affect plant equipment or operating practices and therefore
does not involve a significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Antonio Fernandez, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Branch Chief: David Terao.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: December 15, 2006.
Description of amendment request: The amendment request would
revise the Technical Specifications (TSs) to adopt NRC-approved
Revision 4 to Technical Specification Task Force (TSTF) Standard
Technical Specification Change Traveler TSTF-372, ``Addition of LCO
[Limiting Condition for Operation] 3.0.8, Inoperability of Snubbers.''
The amendment would add (1) a new LCO 3.0.8 addressing when one or more
required snubbers are unable to perform their associated support
function(s) (i.e., the snubber is inoperable) and (2) a reference to
LCO 3.0.8 in LCO 3.0.1 on when LCOs shall be met.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on November 24, 2004 (69 FR 68412), on possible
license amendments adopting TSTF-372 using the NRC's consolidated line
item improvement process (CLIIP) for amending licensee's TSs, which
included a model safety evaluation (SE) and model no significant
hazards consideration (NSHC) determination. 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), which included the resolution of public comments on the
model SE. The May 4, 2005, notice of availability referenced the
November 24, 2004, notice. The licensee has affirmed the applicability
of the following NSHC determination in its application.
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:
[[Page 155]]
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 Accident 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 [a] 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 a Margin of Safety
The proposed change allows a delay time for entering a supported
system TS when the inoperability is due solely to an inoperable
snubber, if risk is assessed and managed. The postulated seismic
event requiring snubbers is a low-probability occurrence and the
overall TS system safety function would still be available for the
vast majority of anticipated challenges. The risk impact of the
proposed TS changes was assessed following the three-tiered approach
recommended in [NRC] 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[,
which is required by the proposed LCO 3.0.8]. 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: Jay Silberg, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action, see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2 (HBRSEP2), Darlington County, South Carolina
Date of application for amendment: April 11, 2006, as supplemented
by letter dated November 9, 2006.
Brief description of amendment: The amendment modifies Technical
Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR),'' to
add a U.S. Nuclear Regulatory Commission-approved topical report to the
listing of analytical methods in TS 5.6.5.b. This change will allow for
the use of the S-RELAP5 thermal-hydraulic analysis code for the non-
loss-of-coolant accident analyses at HBRSEP2.
Date of issuance: November 29, 2006.
Effective date: Effective as of the date of its issuance and shall
be implemented within 60 days.
Amendment No. 211.
Renewed Facility Operating License No. DPR-23. Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: August 29, 2006 (71 FR
51224).
The supplemental letter dated November 9, 2006, provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the scope of the initial
application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 29, 2006.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3), Westchester
County, New York
Date of application for amendment: December 27, 2005, as
supplemented by letter dated August 22, 2006.
Brief description of amendment: The amendment changes consist of
the following changes to the plant Technical Specifications (TSs):
Adoption of Technical Specification Task Force (TSTF)-258,
Revision 4; regarding changes to TS Section 5.0, Administrative
Controls.
Adoption of TSTF-308, Revision 1; regarding the
determination of cumulative and projected dose
[[Page 156]]
contributions in the Radioactive Effluents Control Program (RECP).
Revision of the IP2 definition for dose equivalent iodine-
131 based on NUREG-1431, Revision 3.
Revision of the IP2 RECP requirements based on NUREG-1431,
Revision 3.
Revision of the IP3 Explosive Gas and Storage Tank
Radioactivity Monitoring Program requirements based on NUREG-1431.
Date of issuance: December 13, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: February 14, 2006.
Brief description of amendment: The amendment eliminated the
requirement to verify containment isolation valves that are maintained
locked, sealed, or otherwise secured closed from the monthly position
verification. A new surveillance requirement, (SR) 4.6.1.1.d, was also
added to replace the existing note and reflects the SR for similar
devices located inside containment. In addition, a new note was
included to allow verification by use of administrative means of the
valves and blind flanges that are located in high-radiation areas. In
this regard, the amendment adopts TS Task Force (TSTF) Improved
Standard TS Change Traveler No. 45 (TSTF-45-A), ``Exempt Verification
of Containment Isolation Valves that are Not Locked, Sealed, or
Otherwise Secured.''
Date of issuance: December 18, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 269.
Renewed Facility Operating License No. NPF-6: Amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: April 11, 2006 (71 FR
18373).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 18, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of application for amendment: January 12, 2006.
Brief description of amendment: The amendments revised Technical
Specification 3.6.6, ``Containment Spray and Cooling Systems,''
Surveillance Requirement 3.6.6.3, governing containment cooling train
cooling water flow rate, from `` >2660 gallons per minute (gpm) to each
train'' to `` >2660 gpm to each cooler,'' to accurately reflect the
plant design.
Date of issuance: December 6, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 149, 149, 143 and 143.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications and License.
Date of initial notice in Federal Register: April 25, 2006 (71 FR
23954)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 6, 2006.
No significant hazards consideration comments received: No.
Amendment No.: 250 and 232
Facility Operating License Nos. DPR-26 and DPR-64: The amendment
revised the License and the Technical Specifications.
Date of initial notice in Federal Register: February 14, 2006 (71
FR 7807).
The letter dated August 22, 2006, supplement provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 13, 2006.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of amendment request: June 21, 2006, as supplemented December
12, 2006.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.7.3, ``Intake Cooling Water System,'' Action a, to
increase the allowed outage time for one inoperable intake cooling
water pump from 7 days to 14 days.
Date of issuance: December 12, 2006.
Effective date: December 12, 2006.
Amendment Nos.: 232 and 227.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the TSs.
Date of initial notice in Federal Register: September 12, 2006 (71
FR 53717). The December 12, 2006, Supplement did not affect the
original proposed no significant hazards determination, or expand the
scope of the request as noticed in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 12, 2006.
No significant hazards consideration comments received: No.
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of application for amendment: April 28, 2006.
Brief description of amendment: The amendment modifies the
Technical Specification (TS) requirements for inoperable snubbers by
adding Limiting Condition for Operation 3.0.8. This change is based on
the NRC-approved Technical Specification Task Force (TSTF) standard TS
change TSTF-372, Revision 4. A notice of availability for this TS
improvement using the consolidated line item improvement process was
published in the Federal Register on May 4, 2005 (70 FR 23252).
Date of issuance: December 14, 2006.
Effective date: As of the date of issuance and shall be implemented
within 180 days.
Amendment No.: 263.
Facility Operating License No. DPR-49: The amendment revises the
TSs.
Date of initial notice in Federal Register: (71 FR 43534) August 1,
2006.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 14, 2006.
No significant hazards consideration comments received: No.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2 (NMP2), Oswego County, New York
Date of application for amendment: May 11, 2006.
Brief description of amendment: The amendment revised NMP2
Technical
[[Page 157]]
Specification (TS) 3.1.7, ``Standby Liquid Control (SLC) System,''
(SLCS) by increasing the minimum required NMP2 SLCS pump test discharge
pressure specified in TS Surveillance Requirement 3.1.7.7 from 1235
psig to 1320 psig.
Date of issuance: December 14, 2006.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 117.
Facility Operating License No. NPF-69: Amendment revised the
License and Technical Specifications.
Date of initial notice in Federal Register: September 26, 2006 (71
FR 56192).
The staff's related evaluation of the amendment is contained in a
Safety Evaluation dated December 14, 2006.
No significant hazards consideration comments received: No.
Nuclear Management Company, Docket No. 50-263, Monticello Nuclear
Generating Plant (MNGP), Wright County, Minnesota
Date of application for amendment: September 15, 2005, as
supplemented on April 13, August 21, and August 22, 2006.
Brief description of amendment: The amendment revised the MNGP
licensing basis by implementing the full-scope alternative source term
methodology, resulting in revision of portions of the Technical
Specifications to reflect this licensing basis change.
Date of issuance: December 7, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No: 148.
Facility Operating License No. DPR-22: Amendment revised the
Facility Operating License and Technical Specifications.
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination, and did not expand the scope of the original Federal
Register notice.
Date of initial notice in Federal Register: February 14, 2006 (71
FR 7808).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 7, 2006.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon
Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California
Date of application for amendment: January 13, 2006, as
supplemented by letter dated September 29, 2006.
Brief description of amendment: The amendment revised TS 5.6.5,
``Core Operating Limits Report (COLR),'' by adding Westinghouse Topical
Report WCAP-16009-P-A, ``Realistic Large-Break LOCA [Loss-of-Coolant
Accident] Evaluation Methodology Using the Automated Statistical
Treatment of Uncertainty Method (ASTRUM),'' dated January 2005, as an
approved analytical method for determining the core operating limits
for Diablo Canyon Power Plant, Unit No. 2.
Date of issuance: December 20, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 90 days of issuance.
Amendment No.: 192.
Facility Operating License No. DPR-82: The amendment revised the
Facility Operating License and the Technical Specifications.
Date of initial notice in Federal Register: February 28, 2006 (71
FR 10076).
The September 29, 2006, supplemental letter provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated December 20, 2006.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of application for amendments: March 29, 2006, as supplemented
July 6, 2006.
Brief description of amendments: The amendments revised the
Technical Specifications for containment tendon surveillance to provide
consistency with the requirements of the regulations.
Date of issuance: December 12, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 147, 127.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the licenses and the technical specifications.
Date of initial notice in Federal Register: May 9, 2006 (71 FR
27004). The supplement dated July 6, 2006, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated December 12, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: December 15, 2005, as
supplemented by letters dated June 12 and September 8, 2006 (TS-05-10).
Brief description of amendment: The amendment revises the existing
steam generator tube surveillance program and was modeled after the
U.S. Nuclear Regulatory Commission's approved Technical Specification
Task Force (TSTF) Standard Technical Specification Change Traveler,
TSTF-449, ``Steam Generator Tube Integrity,'' Revision 4. TSTF-449 is
part of the consolidated line item improvement process.
Date of issuance: November 3, 2006.
Effective date: As of the date of issuance and shall be implemented
prior to entering Mode 4 during startup from the Unit 1 Cycle 7
refueling outage.
Amendment No.: 65.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 28, 2006 (71 FR
15489 ). The supplemental letters provided clarifying information that
was within the scope of the initial notice and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated: November 3, 2006.
No significant hazards consideration comments received: No. 65.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: July 6, 2006 (TS-06-04).
Brief description of amendments: The amendment revises Technical
Specification 3.1.3.2, ``Position Indication Systems--Operating,'' for
the Sequoyah Nuclear Plant, Units 1 and 2,
[[Page 158]]
to allow for the use of an alternate means other than movable incore
detectors to monitor the position of a control or shutdown rod should
problems occur with the analog rod position indication system. The use
of this alternate method will reduce the frequency of flux mapping
using movable incore detectors to determine the position of the non-
indicating rod. This will reduce the wear on the movable incore
detector system that is also used to complete other required TS
surveillances.
Date of issuance: December 11, 2006.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 315 and 304.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the technical specifications.
Date of initial notice in Federal Register: August 15, 2006 (71 FR
46938). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: May 1, 2006 (TS-05-10).
Brief description of amendments: The amendments modify the
Technical Specification (TS) Section 6.0, ``Administrative Controls,''
to adopt a Nuclear Regulatory Commission-approved topical report that
extends the burnup limit of the Mark-BW fuel design with M5 alloy.
These amendments also incorporate Technical Specification Task Force
Traveler (TSTF) 363, Revision 0, ``Revised Topical Report References in
Improved Technical Specification 5.6.5, Core Operating Limits Report.''
TSTF-363 makes administrative changes to the format of referenced
topical reports in the TSs.
Date of issuance: November 16, 2006.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos. 314 and 303.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revised the technical specifications.
Date of initial notice in Federal Register: June 20, 2006 (71 FR
35459).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 16, 2006.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: June 16, 2006.
Brief description of amendment: The amendment revises TS Section
5.7.2.11, ``Inservice Testing Program'', consistent with Technical
Specification Task Force (TSTF) Traveler 479, Revision 0, ``Changes to
Reflect Revision of 10 CFR 50.55a'' and TSTF 279, Revision 0, ``Remove
`applicable supports' from Inservice Testing Program.'' The changes
replace references to Section XI of the ASME Boiler and Pressure Vessel
Code with the ASME Operation and Maintenance Code for inservice testing
(IST) activities and removes reference to ``applicable supports'' from
the IST program. In addition, the changes limit the applicability of
Surveillance Requirement 3.0.2 provisions to other normal and
accelerated frequencies specified as two years or less in the IST
program.
Date of issuance: December 18, 2006.
Effective date: As of the date of issuance and shall be implemented
no later than the start of the second 10-year IST interval.
Amendment No. 66.
Facility Operating License No. NPF-90: Amendment revises the
technical specifications.
Date of initial notice in Federal Register: August 15, 2006 (71 FR
46939). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 18, 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 9, 2006.
Brief description of amendment: The amendment revised Technical
Specifications (TS) 1.1, ``Definitions,'' and TS 3.4.16, ``RCS [Reactor
Coolant System] Specific Activity.'' The revisions replaced the current
Limiting Condition for Operation (LCO) 3.4.16 limit on RCS gross-
specific activity with limits on RCS Dose Equivalent I-131 and Dose
Equivalent Xe-133 (DEX). The conditions and required actions for LCO
3.4.16 not being met, as well as surveillance requirements for LCO
3.4.16, are revised. The modes of applicability for LCO 3.4.16 are
extended. The current definition of [Emacr]--Average Disintegration
Energy in TS 1.1 is replaced by the definition of DEX. In addition, the
current definition of Dose Equivalent I-131 in TS 1.1 is revised to
allow alternate NRC-approved thyroid dose conversion factors.
Date of issuance: December 18, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 90 days of the date of issuance.
Amendment No.: 178.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and 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 December 18, 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 9, 2006.
Brief description of amendment: The amendment revised Technical
Specifications (TS) 1.1, ``Definitions,'' and TS 3.4.16, ``RCS [Reactor
Coolant System] Specific Activity.'' The revisions replaced the current
Limiting Condition for Operation (LCO) 3.4.16 limit on RCS gross-
specific activity with limits on RCS Dose Equivalent I-131 and Dose
Equivalent Xe-133 (DEX). The conditions and required actions for LCO
3.4.16 not being met, as well as surveillance requirements for LCO
3.4.16, are revised. The modes of applicability for LCO 3.4.16 are
extended. The current definition of [Emacr]--Average Disintegration
Energy in TS 1.1 is replaced by the definition of DEX. In addition, the
current definition of Dose Equivalent I-131 in TS 1.1 is revised to
allow alternate NRC-approved thyroid dose conversion factors.
Date of issuance: December 18, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 90 days of the date of issuance.
Amendment No.: 178.
Facility Operating License No. NPF-30: The amendment revised the
Operating License and 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 December 18, 2006.
No significant hazards consideration comments received: No.
[[Page 159]]
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: February 1, 2006, as supplemented by
letter dated May 24, 2006.
Brief description of amendment: The amendment revised the Inservice
Testing Program in Section 5.5.8 of the Technical Specifications,
``Administrative Controls, Programs and Manuals,'' to adopt the
Commission-approved Technical Specification Task Force (TSTF)-479,
Revision 0, ``Changes to Reflect Revision of 10CFR50.55a.''
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.: 172.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 28, 2006 (71
FR 10079).
The supplemental letter dated May 24, 2006, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 15, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this December 26, 2006.
For the Nuclear Regulatory Commission.
Timothy J. McGinty,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. E6-22492 Filed 12-29-06; 8:45 am]
BILLING CODE 7590-01-P