[Federal Register Volume 71, Number 108 (Tuesday, June 6, 2006)]
[Notices]
[Pages 32602-32614]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-8450]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from May 12, 2006 to May 24, 2006. The last
biweekly notice was published on May 23, 2006 (71 FR 29671).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
[[Page 32603]]
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: April 26, 2006.
Description of amendment request: The proposed amendment would
modify technical specification (TS) requirements for inoperable
snubbers by adding Limiting Condition for Operation 3.0.8. The changes
are consistent with Nuclear Regulatory Commission approved Industry/
Technical Specification Task Force (TSTF) standard TS change TSTF-372,
Revision 4.
The NRC staff issued a notice of availability of a model safety
evaluation and model no significant hazards consideration (NSHC)
determination for referencing in license amendment applications in the
Federal Register on May 4, 2005 (70 FR 23252). The licensee affirmed
the applicability of the model NSHC determination in its application
dated April 26, 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 proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change allows a delay time before declaring supported
TS systems inoperable when the associated snubber(s) cannot perform its
required safety function. Entrance into Actions or delaying entrance
into Actions is not an initiator of any accident previously evaluated.
Consequently, the probability of an accident previously evaluated
is not significantly increased. The consequences of an accident while
relying on the delay time allowed before declaring a TS supported
system inoperable and taking its Conditions and Required Actions are no
different than the consequences of an accident under the same plant
conditions while relying on the existing TS supported system Conditions
and Required Actions.
[[Page 32604]]
Therefore, the consequences of an accident previously evaluated are
not significantly increased by this change. Therefore, this 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 allows a delay time before declaring supported
TS systems inoperable when the associated snubber(s) cannot perform its
required safety function. The proposed change does not involve a
physical alteration of the plant (no new or different type of equipment
will be installed) or a change in the methods governing normal plant
operations. Thus, this change does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change allows a delay time before declaring supported
TS systems inoperable when the associated snubber(s) cannot perform its
required safety function. The proposed change restores an allowance in
the pre-ISTS conversion TS that was unintentionally eliminated by the
conversion. The pre-ISTS TS were considered to provide an adequate
margin of safety for plant operation, as does the post-ISTS conversion
TS. 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: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Michael L. Marshall, Jr.
Entergy Nuclear Operations, Inc., Docket No. 50-271, Vermont Yankee
Nuclear Power Station (VYNPS), Vernon, Vermont
Date of amendment request: April 22, 2006.
Description of amendment request: The proposed amendment would
relocate the Technical Specification (TS) requirements for shock
suppressors (snubbers) to the Technical Requirements Manual (TRM) and
add a new Limiting Condition for Operation (LCO) 3.0.8.
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 relocate TS 3/4.6.1 to the TRM is
administrative in nature and does not involve the modification of any
plant equipment or affect basic plant operation. Snubber operability
and surveillance requirements will be contained in the TRM to ensure
design assumptions for accident mitigation are maintained.
The proposed change to add LCO 3.0.8 allows a delay time before
declaring supported TS systems inoperable when the associated
snubber(s) cannot perform the required safety function. Entrance into
actions or delaying entrance into actions is not an initiator of any
accident previously evaluated. Consequently, the probability of an
accident previously evaluated is not significantly increased. The
station design and safety analysis assumptions included provisions for
redundancy to provide for periods when redundant systems are out-of-
service per the TS. The proposed snubber LCO ensures that out-of-
service time is minimized and risk is managed per 10 CFR 50.65(a)(4).
Therefore, the consequences of an accident previously evaluated are
not significantly increased by this change.
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 relocate TS 3/4.6.1 to the TRM is
administrative and does not involve any physical alteration of plant
equipment. The proposed change does not change the method by which any
safety-related system performs its function. As such, no new or
different types of equipment will be installed, and the basic operation
of installed equipment is unchanged. The methods governing plant
operation and testing remain consistent with current safety analysis
assumptions.
[* * *]
The proposed change to add LCO 3.0.8 allows a delay time before
declaring supported TS systems inoperable when the associated
snubber(s) cannot perform the required safety function. The proposed
change does not involve a physical alteration of the plant (no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operation.
Therefore, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change to relocate TS 3/4.6.1 to the TRM is
administrative in nature, does not negate any existing requirement, and
does not adversely affect existing plant safety margins or the
reliability of the equipment assumed to operate in the safety analysis.
As such, there are no changes being made to safety analysis
assumptions, safety limits or safety system settings that would
adversely affect plant safety as a result of the proposed change.
Margins of safety are unaffected by requirements that are retained, but
relocated from the TS to the TRM.
[* * *]
The proposed change to add LCO 3.0.8 to TS allows a delay time
before declaring supported TS systems inoperable when the associated
snubber(s) cannot perform the required safety function. The proposed
change retains an allowance in the current VYNPS TS while upgrading it
to be more conservative for snubbers supporting multiple trains or sub-
systems of an associated system. The updated TS will continue to
provide an adequate margin of safety for plant operation upon
incorporation of LCO 3.0.8. The station design and safety analysis
assumptions provide margin in the form of redundancy to account for
periods of time when system capability is reduced. This proposed change
does not reduce that margin.
Therefore, this change does not involve a significant reduction in
a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Travis C. McCullough, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
Branch Chief: Richard Laufer.
[[Page 32605]]
Exelon Generation Company, LLC (EGC), Docket No. 50-374, LaSalle County
Station, Unit 2, LaSalle County, Illinois
Date of amendment request: April 21, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 5.5.13, ``Primary
Containment Leakage Rate Testing Program,'' to reflect a one-time
extension of the LaSalle County Station (LSCS), Unit 2 primary
containment Type A integrated leak rate test (ILRT) date from the
current requirement of no later than December 7, 2008, to prior to
startup following the twelfth LSCS, Unit 2 refueling outage (L2R12).
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?
Response: No.
The proposed changes will revise LSCS, Unit 2, TS 5.5.13, ``Primary
Containment Leakage Rate Testing Program,'' to reflect a one-time
extension of the primary containment Type A Integrated Leak Rate Test
(ILRT) date to ``prior to startup following L2R12.'' The current Type A
ILRT interval of 15 years, based on past performance, would be extended
on a one-time basis by approximately 2% of the current interval.
The function of the primary containment is to isolate and contain
fission products released from the reactor Primary Coolant System (PCS)
following a design basis Loss of Coolant Accident (LOCA) and to confine
the postulated release of radioactive material to within limits. The
test interval associated with Type A ILRTs is not a precursor of any
accident previously evaluated. Type A ILRTs provide assurance that the
LSCS Unit 2 primary containment will not exceed allowable leakage rate
values specified in the TS and will continue to perform their design
function following an accident. The risk assessment of the proposed
changes has concluded that there is an insignificant increase in total
population dose rate and an insignificant increase in the conditional
containment failure probability.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The proposed changes for a one-time extension of the Type A ILRT
for LSCS Unit 2 will not affect the control parameters governing unit
operation or the response of plant equipment to transient and accident
conditions. The proposed changes do not introduce any new equipment,
modes of system operation or failure mechanisms.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the change involve a significant reduction in a margin of
safety?
Response: No.
LSCS Unit 2 is a General Electric BWR/5 plant with a Mark II
primary containment. The Mark II primary containment consists of two
compartments, the drywell and the suppression chamber. The drywell has
the shape of a truncated cone, and is located above the cylindrically
shaped suppression chamber. The primary containment is penetrated by
access, piping and electrical penetrations.
The integrity of the primary containment penetrations and isolation
valves is verified through Type B and Type C local leak rate tests
(LLRTs) and the overall leak tight integrity of the primary containment
is verified by a Type A ILRT, as required by 10 CFR 50, Appendix J,
``Primary Reactor Containment Leakage Testing for Water-Cooled Power
Reactors.'' These tests are performed to verify the essentially leak
tight characteristics of the primary containment at the design basis
accident pressure. The proposed changes for a one-time extension of the
Type A ILRTs do not affect the method for Type A, B or C testing or the
test acceptance criteria.
EGC has conducted a risk assessment to determine the impact of a
change to the LSCS Unit 2 Type A ILRT schedule from a baseline ILRT
frequency of three times in ten years to once in 16.25 years (i.e., 15
years plus 15 months) for the risk measures of Large Early Release
Frequency (i.e., LERF), Total Population Dose, and Conditional
Containment Failure Probability (i.e., CCFP). This assessment indicated
that the proposed LSCS ILRT interval extension has a minimal impact on
public risk.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Daniel S. Collins
R.E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R.E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: May 1, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 1.1, ``Definitions,'' TS 3.4.13,
``RCS [reactor coolant system] Operational Leakage,'' TS 5.5.8, ``Steam
Generator Program,'' and add new specifications (TS 3.4.17) for ``Steam
Generator (SG) Tube Integrity'' and (TS 5.6.7) for ``Steam Generator
Tube Inspection Report.'' The proposed changes are necessary in order
to implement the guidance for the industry initiative on Nuclear Energy
Institute (NEI) 97-06, ``Steam Generator Program Guidelines.''
The NRC staff issued a notice of opportunity for comment in the
Federal Register on March 2, 2005 (70 FR 10298), on possible amendments
adopting Technical Specification Task Force Change Traveller 449,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on May 6, 2005 (70 FR 24126). The
licensee affirmed the applicability of the following NSHC determination
in its application dated May 1, 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
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 an 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,
[[Page 32606]]
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.
An SGTR [steam generator tube rupture] event is one of the design
basis accidents that are analyzed as part of a plant's licensing basis.
In the analysis of a SGTR event, a bounding primary to secondary
LEAKAGE rate equal to the operational LEAKAGE rate limits in the
licensing basis plus the LEAKAGE rate associated with a double-ended
rupture of a single tube is assumed.
For other design basis accidents such as MSLB [main steam line
break], rod ejection, and reactor coolant pump locked rotor the tubes
are assumed to retain their structural integrity (i.e., they are
assumed not to rupture). These analyses typically assume that primary
to secondary LEAKAGE for all SGs is 1 gallon per minute or increases to
1 gallon per minute as a result of accident induced stresses. The
accident induced leakage criterion introduced by the proposed changes
accounts for tubes that may leak during design basis accidents. The
accident induced leakage criterion limits this leakage to no more than
the value assumed in the accident analysis.
The SG performance criteria proposed change 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 I-131 in the primary coolant and the primary to
secondary LEAKAGE rates resulting from an accident. Therefore, limits
are included in the plant technical specifications for operational
leakage and for DOSE EQUIVALENT I-131 in primary coolant to ensure the
plant is operated within its analyzed condition. The typical analysis
of the limiting design basis accident assumes that primary to secondary
leak rate after the accident is 1 gallon per minute with no more than
[500 gallons per day or 720 gallons per day] in any one SG, and that
the reactor coolant activity levels of DOSE EQUIVALENT I-131 are at the
TS values before the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TSs and enhances the requirements
for SG inspections. The proposed change does not adversely impact any
other previously evaluated design basis accident and is an improvement
over the current TSs.
Therefore, the proposed change does not affect the consequences of
a SGTR accident and the probability of such an accident is reduced. In
addition, the proposed changes do not affect the consequences of an
MSLB, rod ejection, or a reactor coolant pump locked rotor event, or
other previously evaluated accident.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The proposed performance based requirements are an improvement over
the requirements imposed by the current technical specifications.
Implementation of the proposed SG Program will not introduce any
adverse changes to the plant design basis or postulated accidents
resulting from potential tube degradation. The result of the
implementation of the SG Program will be an enhancement of SG tube
performance. Primary to secondary LEAKAGE that may be experienced
during all plant conditions will be monitored to ensure it remains
within current accident analysis assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility of a
new or different [kind] of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in [a] Margin of Safety
The SG tubes in pressurized water reactors are an integral part of
the reactor coolant pressure boundary and, as such, are relied upon to
maintain the primary system's pressure and inventory. As part of the
reactor coolant pressure boundary, the SG tubes are unique in that they
are also relied upon as a heat transfer surface between the primary and
secondary systems such that residual heat can be removed from the
primary system. In addition, the SG tubes isolate the radioactive
fission products in the primary coolant from the secondary system. In
summary, the safety function of an SG is maintained by ensuring the
integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube inspection,
assessment, repair, and plugging. The requirements established by the
SG Program are consistent with those in the applicable design codes and
standards and are an improvement over the requirements in the current
TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the TS.
The NRC staff proposes to determine that the amendments request
involves no significant hazards consideration.
Attorney for licensee: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Branch Chief: Richard J. Laufer.
Southern 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 amendment requests: April 28, 2006.
Description of amendment requests: The proposed change will
increase the minimum allowed boron concentration of the spent fuel pool
and allow credit for soluble boron, guide tube inserts (GT-Inserts)
made from borated stainless steel, and fuel storage patterns in place
of Boraflex.
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.
[[Page 32607]]
Dropped Fuel Assembly
There is no significant increase in the probability of a fuel
assembly drop accident in the spent fuel pool when assuming a complete
loss of the Boraflex panels in the spent fuel pool racks and
considering the presence of soluble boron in the spent fuel pool water
for criticality control.
Neither the presence of soluble boron in the spent fuel pool water,
nor the placement of borated stainless steel guide tube inserts (GT-
Inserts) in the fuel assemblies for criticality control, will increase
the probability of a fuel assembly drop accident. The handling of the
fuel assemblies in the spent fuel pool has always been performed in
borated water, and the quantity of Boraflex remaining in the racks or
GT-Inserts placed in the fuel assemblies, has no affect on the
probability of such a drop accident.
Southern California Edison (SCE) has performed a criticality
analysis which shows that the consequences of a fuel assembly drop
accident in the spent fuel pool are not affected when considering a
complete loss of the Boraflex in the spent fuel racks and the presence
of soluble boron. The rack Keff remains less than or equal
to 0.95.
The fuel, the fuel rack, and the fuel pool qualifications have been
evaluated and determined to be unaffected by the installation of the
GT-Inserts. The mechanical design configuration of the GT-Inserts is
similar to the shape, size, and weight of a control element assembly
(CEA) finger. Each of the GT-Inserts are approximately 0.78 inch
outside diameter (OD) solid stainless steel, with a boron content of
approximately 2 weight percent (w/o). A small counterbore is machined
at the top for handling and a rounded bottom is machined. The OD of
these GT-Inserts is less than that of a CEA finger. The material
(borated stainless steel) is American Society for Testing and Materials
(ASTM) approved and has been licensed by the United States Nuclear
Regulatory Commission (NRC) for use in spent fuel storage technologies
and spent fuel pools. The structural effect of the weight of the GT-
Inserts on the fuel, the fuel rack, and the fuel pool structural
interfaces and drop qualifications are unaffected. This is because the
addition of five GT-Inserts (which increases the dry weight of a fuel
assembly by 110 lbs.) brings the total weight to 1551 lbs. which is
enveloped by the 2904 lbs. assumed in the calculation for fuel rack
design.
Fuel Misloading
There is no significant increase in the probability of the
accidental misloading of spent fuel assemblies into the spent fuel
racks when assuming a complete loss of the Boraflex panels and
considering the presence of soluble boron in the pool water for
criticality control. Fuel assembly placement will continue to be
controlled pursuant to approved fuel handling procedures and will be in
accordance with Technical Specification (TS) 3.7.18[,] ``Spent Fuel
Assembly Storage[,]'' and Licensee Controlled Specification (LCS)
4.0.100, ``Fuel Storage Patterns,'' which will specify spent fuel rack
storage configuration limitations.
There is no increase in the consequences of the accidental
misloading of a spent fuel assembly into the spent fuel racks. The
criticality analysis, performed by SCE, demonstrates that the pool
Keff will be maintained less than or equal to 0.95 following
an accidental misloading by the boron concentration of the pool. The
proposed TS 3.7.17[,] ``Fuel Storage Pool Boron Concentration[,]'' will
ensure that an adequate spent fuel pool boron concentration is
maintained.
Change in Spent Fuel Temperature
There is no significant increase in the probability of either the
loss of normal cooling to the spent fuel pool water or a decrease in
pool water temperature from a large emergency makeup when assuming a
complete loss of the Boraflex panels and considering the presence of
soluble boron in the spent fuel pool water. A high proposed
concentration (>2000 parts per million (ppm)) of soluble boron is
consistent with current operating practices maintained in the spent
fuel pool water. The proposed minimum boron concentration of 2000 ppm
in TS 3.7.17 will ensure that an adequate concentration is maintained
in the spent fuel pools.
A loss of normal cooling to the spent fuel pool water causes an
increase in the temperature of the water passing through the stored
fuel assemblies. This causes a decrease in the water density, and when
coupled with the assumption of a complete loss of Boraflex, may result
in a positive reactivity addition. However, the additional negative
reactivity provided by the boron concentration limit in the proposed TS
3.7.17 will compensate for the increased reactivity which could result
from a loss of spent fuel pool cooling. Because adequate soluble boron
will be maintained in the spent fuel pool water to maintain
Keff less than or equal to 0.95, the consequences of a loss
of normal cooling to the spent fuel pool will not be increased.
The thermal considerations of the fuel are unaffected by the
presence of the GT-Inserts because the guide tube is designed for the
presence of a CEA; therefore, it is not a primary coolant flow area.
The fuel rack normal thermal cooling and malfunctioned blocked cooling
scenarios are unaffected by the presence of the GT-Inserts in the fuel
assemblies.
The proposed change does not involve an 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 consideration of criticality accidents in the spent fuel pool
are not new or different. They have been analyzed in the Updated Final
Safety Analysis Report (UFSAR) and in previous submittals to the NRC.
Specific accidents considered and evaluated include fuel assembly drop,
fuel assembly misloading in the racks, and spent fuel pool water
temperature changes.
The possibility for creating a new or different kind of accident is
not credible. Neither Boraflex [n]or soluble boron are accident
initiators. The proposed change takes credit for soluble boron in the
spent fuel pool while maintaining the necessary margin of safety.
Because soluble boron has always been present in the spent fuel pool, a
dilution of the spent fuel pool soluble boron has always been a
possibility. However, a criticality accident resulting from a dilution
accident was not considered credible. For this proposed amendment, SCE
performed a spent fuel pool dilution analysis, which demonstrated that
a dilution of the boron concentration in the spent fuel pool water
which could increase the rack Keff to greater than 0.95
(constituting a reduction of the required margin to criticality) is not
a credible event. The requirement to maintain boron concentration in
the spent fuel pool water for reactivity control will have no effect on
normal pool operations and maintenance. There are no changes in
equipment design or plant configuration.
The possibility of accidentally withdrawing a GT-Insert is
minimized because special tooling is required to remove it, and it is
completely contained within the guide tubes of the designated
assemblies. Potential misloading of the GT-Inserts is minimized due to
the design of the
[[Page 32608]]
installation equipment, procedural controls, and double verification
that will be in place to ensure the GT-Inserts are installed properly.
The possibility of accidentally withdrawing a CEA is minimized
because specialized tooling is required for withdrawing a CEA from a
fuel assembly. It is physically possible for the spent fuel handling
tool to bind on a CEA after ungrappling from a fuel assembly and
raising the tool. However, existing SONGS [San Onofre Nuclear
Generating Station] procedures require that the operator validate
``tool weight only'' on the spent fuel handling machine's load cell
read out after ungrappling from a fuel assembly and raising the hoist
slightly, and to report this information to the engineer directing the
fuel movement.
Therefore, the proposed change will not result in 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 TS changes proposed by this license amendment request and the
resulting spent fuel storage operation limits will provide adequate
safety margin to ensure that the stored fuel assembly array will always
remain subcritical. Those limits are based on a San Onofre Nuclear
Generating Station (SONGS) Units 2 and 3 plant specific analysis that
was performed in accordance with a methodology previously approved by
the NRC.
The proposed change takes partial credit for soluble boron in the
spent fuel pool. SCE's analyses show that spent fuel storage
requirements meet the following NRC acceptance criteria for preventing
criticality outside the reactor.
(1) The neutron multiplication factor, Keff, including
all uncertainties, shall be less than 1.0 when flooded with unborated
water, and
(2) The neutron multiplication factor, Keff, including
all uncertainties, shall be less than or equal to 0.95 when flooded
with borated water.
The criticality analysis utilized credit for soluble boron to
ensure Keff will be less than or equal to 0.95 under normal
circumstances, and storage configurations have been defined using a 95/
95 Keff calculation to ensure that the spent fuel rack will
be less than 1.0 with no soluble boron. Soluble boron credit is used to
provide safety margin by maintaining Keff less than or equal
to 0.95 including uncertainties, tolerances[,] and accident conditions
in the presence of spent fuel pool soluble boron. SCE evaluated the
loss of a substantial amount of soluble boron from the spent fuel pool
water which could lead to Keff exceeding 0.95 and showed
that it was not credible.
Also, the spent fuel rack Keff will remain less than 1.0
with the spent fuel pool flooded with unborated water.
Decay heat, radiological effects, and seismic loads are unchanged
by the absence of Boraflex.
The mechanical properties and the weight of the fuel assemblies
remain essentially unchanged with the inclusion of the weight of five
GT-Inserts per assembly. The original mechanical and thermal analysis
of the fuel assembly/fuel rack and fuel pool building interfaces
currently approved remain valid and conservative.
Therefore, the proposed change does not involve a significant
reduction in the plant's 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 requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Branch Chief: David Terao.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 30, 2006.
Description of amendment request: The proposed amendments revise
Technical Specification 3.3.3.6, ``Accident Monitoring
Instrumentation,'' with respect to the required action for inoperable
Wide Range Reactor Coolant Temperature, Wide Range Steam Generator
Water Level, and Auxiliary Feedwater (AFW) Flow.
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 increase in the allowed outage times for the Reactor
Coolant Outlet Temperature--Wide Range, Reactor Coolant Inlet
Temperature--Wide Range, Steam Generator [Water] Level--Wide Range, and
the AFW Flow does not involve a significant increase in the probability
of an accident previously evaluated because these are accident
monitoring functions that have no effect on the potential for accident
initiation. The proposed deletion of the existing requirements in
ACTION 38 is an administrative change. Since these requirements are not
currently applied to any plant equipment, this change cannot affect the
probability of any accident previously evaluated.
The proposed increase in the allowed outage times for the Reactor
Coolant Outlet Temperature--Wide Range, Reactor Coolant Inlet
Temperature--Wide Range, Steam Generator [Water] Level--Wide Range, and
AFW Flow does not involve a significant increase in the consequences of
an accident previously evaluated because the availability of redundant
and diverse indications provides adequate assurance that the operator
will be able to determine the post-accident status of the secondary
heat sink.
The proposed deletion of the existing requirements in ACTION 38 is
an administrative change. Since these requirements are not currently
applied to any plant equipment, this change cannot affect the
consequence 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 increase in the allowed outage times for the Reactor
Coolant Outlet Temperature--Wide Range, Reactor Coolant Inlet
Temperature--Wide Range, Steam Generator [Water] Level--Wide Range, and
the AFW Flow does not create the possibility of a new or different kind
accident from any accident previously evaluated because the proposed
change affects only the allowed outage time for accident monitoring
instrumentation and involves no changes to plant design, plant
configuration or operating procedures.
The proposed deletion of the existing requirements in ACTION 38 is
an administrative change. Since these requirements are not currently
applied to any plant equipment, this change cannot create the
possibility of any kind of accident.
(3) Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed increase in the allowed outage times for the Reactor
Coolant Outlet Temperature--Wide Range, Reactor Coolant Inlet
Temperature--
[[Page 32609]]
Wide Range, Steam Generator [Water] Level--Wide Range, and AFW Flow
does not involve a significant reduction in the margin of safety
because the availability of redundant and diverse indications provides
adequate assurance that the operator will be able to determine the
post-accident status of the secondary heat sink.
The proposed deletion of the existing requirements in ACTION 38 is
an administrative change. Since these requirements are not currently
applied to any plant equipment, this change cannot affect the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: David Terao.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: February 21, 2006.
Brief description of amendments: The amendments revise Technical
Specification (TS) 5.6.5 entitled, ``Core Operating Limits Report
(COLR),'' to revise the listed Loss-of-Coolant Accident (LOCA) and non-
LOCA analysis methodologies used at Comanche Peak Steam Electric
Station, Units 1 and 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change involves an administrative change only.
Designation of the accident analysis methodologies, described in ERX-
04-004 and ERX-04-005, as approved analytical methods is required to
maintain the accuracy of the Technical Specification 5.6.5 (Core
Operating Limits Report) and to maintain consistency with the
resolution of issues as prescribed in 10 CFR 50.46. Therefore, the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change involves an administrative change only.
Technical Specification 5.6.5 is being changed to reference the revised
accident analysis methodologies currently under NRC review. No actual
plant equipment will be affected by the proposed change. Therefore, the
proposed change does not create the possibility of a new or different
kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
Margin of safety is associated with the confidence in the ability
of the fission product barriers (i.e., fuel and fuel cladding, Reactor
Coolant System pressure boundary, and containment structure) to limit
the level of radiation dose to the public. This request involves an
administrative change (subject to NRC approval) only to incorporate the
NRC-approved methodologies into the allowable analysis methodologies
specified in Technical Specification 5.6.5. No actual plant equipment
will be affected by the proposed change. The compliance of the revised
methodology with the requirements of 10 CFR 50.46 and Appendix K will
be addressed through the NRC staff's review of the topical reports.
Therefore, it is concluded that the use of the proposed methodology
will not degrade the confidence in the ability of the fission product
barriers to limit the level of radiation dose to the public. Therefore
the proposed change does not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: David Terao.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: February 21, 2006.
Brief description of amendments: The amendments would revise
Technical Specifications (TS) 3.3.1, 3.3.2, 3.4.5, 3.4.6, and 3.4.7,
``Reactor Trip System (RTS) Instrumentation,'' ``Engineered Safety
Feature System Actuation (ESFAS) Instrumentation,'' ``RCS [Reactor
Coolant System] Mode 3,'' ``RCS Loops-Mode 4,'' and ``RCS Loops-Mode 5,
Loops Filled,'' respectively. The revisions reflect the different steam
generator water level trip setpoints and steam generator inventory
requirements associated with the planned replacement of the steam
generators in Comanche Peak Steam Electric Station, Unit 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed TS changes affect the protective and mitigative
capabilities of the plant; none of the changes impact the initiation or
probability of occurrence of any accident.
The consequences of accidents evaluated in the FSAR [Final Safety
Analysis Report] that could be affected by this proposed change are
those in which the steam generator water level trip functions are
credited for initiating a protective or mitigative function. These
transients and accidents have been analyzed and all relevant event
acceptance criteria were shown to be satisfied. The radiological dose
consequences are unaffected. Therefore, there is no increase in the
consequences of an accident previously evaluated.
The actual proposed setpoint values were determined using an
uncertainty methodology previously approved by the NRC for this
application. These values provide adequate assurance that required
protective and mitigative functions will be initiated as assumed in the
transient and accident analyses. Therefore, there is no increase in the
consequences of an accident previously evaluated.
The proposed revisions to the [Delta]76 steam generator inventory,
required to ensure that the steam generators can provide an effective
heat sink, are consistent with the current design requirements.
Therefore, the proposed changes do not involve a significant increase
in the probability or consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of
[[Page 32610]]
accident from any accident previously evaluated?
Response: No.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result of
these changes. There will be no adverse effect or challenges imposed on
any safety-related system as a result of these changes. There are no
changes which would cause the malfunction of safety-related equipment,
assumed to be operable in the accident analyses, as a result of the
proposed Technical Specification changes. No new equipment performance
burdens are imposed. The possibility of a new or different malfunction
of safety-related equipment is not created. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes to the Steam Generator Water Level-Low-Low and
Steam Generator Water Level-High-High trip function setpoints protect
the assumed safety analysis limits established in the transient and
accident analyses. When used in the transient and accident analyses,
all relevant event acceptance criteria are satisfied. Therefore, these
proposed changes do not result in the reduction in a margin of safety.
The proposed changes to the [Delta]76 steam generator inventory
requirements, which ensure the steam generators can function as an
effective heat sink during required shutdown operating modes, are
consistent with the existing design and licensing bases. Therefore,
these proposed changes do not result in the reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: David Terao.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Georgia Power Company, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia
Date of amendment request: March 17, 2006.
Brief description of amendment request: The proposed amendment
would add a license condition to Section 2.C of the Edwin I. Hatch
Nuclear Plant, Unit Nos. 1 and 2, Operating Licenses. This license
condition will authorize the licensee to credit administering potassium
iodide (KI) to reduce the 30-day post-accident thyroid radiological
dose to the operators in the main control room for an interim period of
approximately 4 years. In addition, the design-basis accident analysis
section of the Updated Final Safety Analysis Reports will be updated to
reflect crediting of KI.
Date of publication of individual notice in Federal Register: March
27, 2006 (71 FR 15223).
Expiration date of individual notice: 30-day date April 26, 2006;
60-day date May 26, 2006.
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].
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: December 19, 2005.
Brief description of amendment: The amendment revised the Technical
Specification (TS) to make permanent the temporary changes to TS Table
3.3.8.1-1 previously approved by Amendment No. 147. TS Table 3.3.8.1-1
is revised to delete the temporary note, correct the number of Required
Channels per Division for the Loss of Power (LOP) time delay functions,
and delete the requirement to perform Surveillance Requirement
3.3.8.1.2, the monthly Channel Functional Test, on certain LOP time
delay functions.
Date of issuance: May 17, 2006.
Effective date: As of the date of issuance and shall be implemented
prior to expiration of the temporary change on June 1, 2006, provided
by Amendment No. 147.
Amendment No.: 151.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specfications.
[[Page 32611]]
Date of initial notice in Federal Register: March 14, 2006 (71 FR
13173).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 17, 2006.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket No. 50-
278, Peach Bottom Atomic Power Station, Unit 3, York and Lancaster
Counties, Pennsylvania
Date of application for amendment: July 6, 2005, as supplemented
March 15 and April 7, 2006.
Brief description of amendments: The proposed changes extend the
use of the Peach Bottom Atomic Power Station, Unit 3, pressure-
temperature (P-T) limits specified in the Technical Specifications
(TSs) from 22 to 32 effective full-power years.
Date of issuance: May 12, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 263.
Renewed Facility Operating License No. DPR-56: The amendment
revised the TSs.
Date of initial notice in Federal Register: August 2, 2005 (70 FR
44402). The supplements dated March 15, 2006, and April 7, 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
amendment is contained in a Safety Evaluation dated May 12, 2006.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: October 21, 2005, as
supplemented February 28, March 28 and April 24, 2006.
Brief description of amendment: The amendment revised the Operating
License and Technical Specifications to allow operation of St. Lucie
Unit 2 with a reduced reactor coolant system flow rate of 300,000 gpm
and a reduction in the maximum thermal power to 89 percent of the rated
thermal power. The flow rate of 300,000 gpm conservatively bounds an
analyzed steam generator tube plugging level of 42 percent per steam
generator.
Date of Issuance: May 16, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 145.
Renewed Facility Operating License No. NPF-16: Amendment revised
the TS.
Date of initial notice in Federal Register: December 20, 2005 (70
FR 75492). The February 28, March 28 and April 24, 2006, supplements
did not affect the original proposed no significant hazards
determination, or expand the scope of the request as noticed in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 16, 2006.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: September 22, 2005, as supplemented by
letters dated March 24, 2006, and April 28, 2006.
Description of amendment request: The proposed amendment revised
the Seabrook Station, Unit No. 1 Technical Specifications (TSS) to
increase the licensed thermal power level by 1.7% to 3648 megawatts
thermal.
Date of issuance: May 22, 2006.
Effective date: As of its date of issuance, and shall be
implemented within 12 months.
Amendment No.: 110.
Facility Operating License No. NPF-86: The amendment revised the
Tss and the License.
Date of initial notice in Federal Register: November 8, 2005 (70 FR
67748). The licensee's letters dated March 24, 2006, and April 28,
2006, provided clarifying information that did not change the scope of
the proposed amendment as described in the original notice of proposed
action published in the Federal Register, and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 22, 2006.
No significant hazards consideration comments received: No.
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 application for amendments: July 29, 2005.
Brief description of amendments: The amendments revised Technical
Specification 3.7.5, ``Auxiliary Feedwater (AFW) System,'' to change
the frequency of Surveillance Requirement 3.7.5.6 from 92 days to 24
months.
Date of issuance: May 17, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 120 days of issuance.
Amendment Nos.: 186 and 188.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 11, 2005 (70 FR
59086).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 17, 2006.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: November 29, 2005.
Brief description of amendment: This amendment for V. C. Summer
revises TSs by eliminating the requirements to submit monthly operating
reports and certain annual reports.
Date of issuance: May 19, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 175.
Renewed Facility Operating License No. NPF-12: Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: March 14, 2006 (71 FR
13178).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 19, 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 13, 2005.
Brief description of amendment: The amendment changes the steam
generator (SG) level requirement for Limiting Condition for Operation
3.4.7.b and Surveillance Requirements 3.4.5.2, 3.4.6.3 and 3.4.7.2 from
greater than or equal (>=) to 6 percent (%) to >= 32% following
replacement of the SGs during the Unit 1, Cycle 7 refueling outage.
Date of issuance: May 5, 2006.
Effective date: As of the date of issuance and shall be implemented
prior to entering Mode 5 upon restart
[[Page 32612]]
from the Unit 1 Cycle 7 (U1C7) Refueling Outage.
Amendment No.: 61.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 14, 2006 (71
FR 7814).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 5, 2006.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: March 8, 2005.
Brief description of amendments: These amendments revised the
auxiliary feedwater (AFW) requirements of Technical Specifications
(TSs) 3.6, ``Turbine Cycle,'' and 4.8, ``Auxiliary Feedwater System,''
to eliminate the inconsistency between the AFW pump requirements and
the required actions, establish consistency with the Improved TSs, and
add an AFW flowpath allowed outage time along with required actions.
Date of issuance: February 23, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 246 and 245.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments change the Technical Specifications.
Date of initial notice in Federal Register: April 26, 2005 (70 FR
21465).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 23, 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 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
[[Page 32613]]
are problems in accessing the document, contact the PDR Reference staff
at 1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------
\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, 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).
Southern California Edison Company, et al., Docket No. 50-362, San
Onofre Nuclear Generating Station, Unit 3, San Diego County, California
Date of amendment request: May 4, 2006.
Description of amendment request: Allowed repairing a line in the
shutdown cooling (SDC) system with the unit in Mode 4. This repair plan
caused Unit 3 to be out of compliance with the licensing basis of the
SDC system for the limited duration of the repair, but not to exceed 7
days.
Date of issuance: May 5, 2006.
Effective date: Immediate.
Amendment No.: 194.
Facility Operating License No. (NPF-15): Amendment revised the
Updated Final Safety Analysis Report, Section 5.4.7.1.2.C. with a note
that states that the change is only applicable from the date of
issuance of the amendment until the repair is completed on the SDC line
or 7 days, whichever occurs first.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No. The Commission's related evaluation of the
amendment, finding of emergency circumstances, state consultation, and
final NSHC determination are contained in a safety evaluation dated May
5, 2006.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
[[Page 32614]]
NRC Branch Chief: David Terao.
Dated at Rockville, Maryland, this 25th day of May 2006.
For the Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E6-8450 Filed 6-5-06; 8:45 am]
BILLING CODE 7590-01-P