[Federal Register Volume 71, Number 147 (Tuesday, August 1, 2006)]
[Notices]
[Pages 43528-43543]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-6597]
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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 July 7, 2006 to July 19, 2006. The last
biweekly notice was published on July 18, 2006 (71 FR 40742).
[[Page 43529]]
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
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of
[[Page 43530]]
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].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: September 29, 2005, as supplemented by
letter dated July 5, 2006.
Description of amendments request: The amendments revised the
Physical Security Plan to clarify the description of the owner
controlled area vehicle checkpoint.
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 amendment, which will clarify the description of a
security feature of the Owner Controlled Area (OCA) Checkpoint, does
not reduce the ability of the Security organization to prevent
radiological sabotage and, therefore, does not increase the
probability or consequences of a radiological release previously
evaluated. The proposed Security Plan changes will not affect any
important to safety systems or components, their mode of operation
or operating strategies. The proposed Security Plan changes have no
affect on accident initiators or mitigation. Therefore, the proposed
amendment to the Security Plan will 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 amendment to clarify the description of a security
feature of the OCA Checkpoint does not affect the operation of
systems important to safety. The Security Plan amendment does not
affect any of the parameters or conditions that could contribute to
the initiation of any accident. No new accident scenarios are
created as a result of the proposed Security Plan changes. In
addition, the design functions of equipment important to safety are
not altered as a result of the proposed Security Plan changes.
Therefore, the proposed Security Plan changes will not create the
possibility of a new or different accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed Security Plan changes will not affect any important
to safety systems or components, their mode of operation, or
operating strategies. The proposed Security Plan changes have no
affect on accident initiators or mitigation. The proposed amendment
to the Security Plan does not reduce the effectiveness of any
security/safeguards measures currently in place. Therefore, the
proposed Security Plan changes will not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three 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: Janet S. Mueller, Director, Law Department,
Arizona Public Service Company, P.O. Box 52034, Mail Station 8695,
Phoenix, Arizona 85072-2034.
NRC Branch Chief: David Terao.
Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of amendment request: June 28, 2006.
Description of amendment request: The proposed amendment changed
Kewaunee Power Station (KPS) Technical Specifications 3.3.b.3.B and
3.3.b.4.A to increase the minimum required boron concentration in the
refueling water storage tank (RWST) from 2400 parts per million (ppm)
to 2500 ppm.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Increasing the minimum required boron concentration in the RWST
does not add, delete, or modify any KPS systems, structures, or
components (SSCs). The RWST and its contents are not accident
initiators. Rather, they are designed for accident mitigation. The
effects of an increase in the minimum RWST boron concentration from
2400 ppm to 2500 ppm are bounded by existing evaluations and
determined to be acceptable. Thus, the proposed increase in minimum
RWST boron concentration has no adverse effect on the ability of the
plant to mitigate the effects of design basis accidents.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
Increasing the minimum required boron concentration in the RWST
does not change
[[Page 43531]]
the design function of the RWST or the SSCs designed to deliver
borated water in the RWST to the [reactor] core. Increasing the
minimum required boron concentration in the RWST does not create any
credible new failure mechanisms or malfunctions for plant equipment
or the nuclear fuel. The safety function of the borated water in the
RWST is not being changed.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
An evaluation has been performed showing that maintaining RWST
boron concentration above 2500 ppm continues to assure acceptable
results for design basis accident analyses [ ] considering the
reactivity of the core. Increasing the minimum boron concentration
in the RWST from 2400 ppm to 2500 ppm increases the margin of safety
in the KPS safety analyses, since additional post-accident negative
reactivity will be available to the core. This additional negative
reactivity more than compensates for the additional reactivity in
the core due to the unanticipated prolonged shutdown periods in
Cycle 27. Additionally, the proposed new minimum boron concentration
of 2500 ppm is within the range required by current safety analyses
(i.e., 2400 ppm to 2625 ppm), and well below the currently
acceptable maximum boron concentration of 2625 ppm.
The proposed amendment does not result in altering or exceeding
a design basis or safety limit for the plant. All current fuel
design criteria will continue to be satisfied, and the safety
analyses of record (except for the postLOCA sump boron
concentration), including evaluations of the radiological
consequences of design basis accidents, will remain applicable.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Branch Chief: L. Raghavan.
Entergy Nuclear Operations, Inc., Docket Nos. 50-247 and 50-286, Indian
Point Nuclear Generating Unit Nos. 2 and 3, Westchester County, New
York
Date of amendment request: May 31, 2006.
Description of amendment request: The proposed amendment revised
the Technical Specification (TS) requirements related to steam
generator (SG) tube integrity. Specifically, it would revise the TS
definition of LEAKAGE; TS 3.4.13, ``Reactor Coolant System (RCS)
Operational Leakage;'' TS 5.5.7 (Indian Point Unit 2) and TS 5.5.8
(Indian Point Unit 3), ``Steam Generator (SG) Program;'' TS 5.6.7
(Indian Point Unit 2) and TS 5.6.8 (Indian Point Unit 3), ``SG Tube
Inspection Report;'' and would create new TS 3.4.17, ``SG Tube
Integrity.''
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF 449, Revision 4. The NRC staff
issued a notice of opportunity for comment in the Federal Register on
March 2, 2005 (70 FR 10298), on possible amendments concerning TSTF-
449, including a model safety evaluation and model no significant
hazards consideration (NSHC) determination, using the consolidated line
item improvement process (CLIIP). 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 31, 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, 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 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 Nuclear Energy Institute (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 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 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 a main steam line break (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.
[[Page 43532]]
Primary to secondary LEAKAGE that may be experienced during all
plant conditions will be monitored to ensure it remains within
current accident analysis assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the SG Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the SG Program are consistent with those in the
applicable design codes and standards and are an improvement over
the requirements in the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Richard J. Laufer.
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: April 27, 2006.
Description of amendment request: The proposed amendments revised
the Technical Specifications (TSs) relating to Steam Generator (SG)
inspection. Specifically, TS 3/4.4.5, Surveillance Requirements, and TS
3/4.4.6, Reactor Coolant System Leakage, would be modified to clearly
delineate the scope of the inservice inspections required in the tube
sheet regions of the SGs.
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?
Of the various accidents previously evaluated, the proposed
changes only affect the SG tube rupture (SGTR) event evaluation and
the postulated steam line break [SLB] accident evaluation. Loss-of-
coolant accident (LOCA) conditions cause a compressive axial load to
act on the tube. Therefore, since the LOCA tends to force the tube
into the tubesheet rather than pull it out, it is not a factor in
this amendment request. Another faulted load consideration is a safe
shutdown earthquake (SSE); however, the seismic analysis of Series
44F SGs has shown that axial loading of the tubes is negligible
during a SSE.
For the SGTR event, the required structural margins of the SG
tubes will be maintained by the presence of the tubesheet. Tube
rupture is precluded for cracks in the hydraulic expansion region
due to the constraint provided by the tubesheet. Therefore,
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR
[Pressurized-Water Reactor] Steam Generator Tubes,'' margins against
burst are maintained for both normal and postulated accident
conditions.
The limited inspection length of 17 inches supplies the
necessary resistive force to preclude pullout loads under both
normal operating and accident conditions. The contact pressure
results from the hydraulic expansion process, thermal expansion
mismatch between the tube and tubesheet and from the differential
pressure between the primary and secondary side. The proposed
changes do not affect other systems, structures, components or
operational features. Therefore, the proposed change results in no
significant increase in the probability of the occurrence of a SGTR
event.
The consequences of an SGTR event are affected by the primary-
to-secondary leakage flow during the event. Primary-to-secondary
leakage flow through a postulated broken tube is not affected by the
proposed change since the tubesheet enhances the tube integrity in
the region of the hydraulic expansion by precluding tube deformation
beyond its initial expanded outside diameter. The resistance to both
tube rupture and collapse is strengthened by the tubesheet in that
region. At normal operating pressures, leakage from primary water
stress corrosion cracking (PWSCC) below 17 inches from the top of
the tubesheet is limited by both the tube-to-tubesheet crevice and
the limited crack opening permitted by the tubesheet constraint.
Consequently, negligible normal operating leakage is expected from
cracks within the tubesheet region.
The probability of a SLB is unaffected by the potential failure
of a SG tube as the failure of a tube is not an initiator for a SLB
event. SLB leakage is limited by leakage flow restrictions resulting
from the crack and tube-to-tubesheet contact pressures that provide
a restricted leakage path above the indications and also limit the
degree of crack face opening compared to free span indications. The
leak rate during postulated accident conditions would be expected to
be less than twice that during normal operation for indications near
the bottom of the tubesheet (including indications in the tube end
welds) based on the observation that while the driving pressure
increases by about a factor of two, the flow resistance increases
with an increase in the tube-to-tubesheet contact. While such a
decrease is rationally expected, the postulated accident leak rate
is bounded by twice the normal operating leak rate if the increase
in contact pressure is ignored. Since normal operating leakage is
limited to less than 150 gpd, the attendant accident condition leak
rate, assuming all leakage to be from lower tubesheet indications,
would be bounded by 300 gpd. This value is less than the 500 gpd
leak rate assumed during a postulated SLB in the Turkey Point Units
3 and 4 Updated Final Safety Analysis Report (UFSAR).
Therefore, based on the above evaluation, 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?
The proposed changes do not introduce any changes or mechanisms
that create the possibility of a new or different kind of accident.
Tube bundle integrity is expected to be maintained for all plant
conditions upon implementation of the limited tubesheet inspection
depth methodology. The proposed changes do not introduce any new
equipment or any change to existing equipment. No new effects on
existing equipment are created nor are any new malfunctions
introduced.
Therefore, based on the above evaluation, the proposed changes
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes maintain the required structural margins of
the SG tubes for both normal and accident conditions. NEI [Nuclear
Energy Institute] 97-06, Rev. 2 and RG 1.121 are used as the basis
in the development of the limited tubesheet inspection depth
methodology for determining that SG tube integrity considerations
are maintained within acceptable limits. RG 1.121 describes a
[[Page 43533]]
method acceptable to the NRC staff for meeting General Design
Criteria 14, 15, 31, and 32 by reducing the probability and
consequences of an SGTR. RG 1.121 concludes that by determining the
limiting safe conditions of tube wall degradation beyond which tubes
with unacceptable cracking, as established by inservice inspection,
should be removed from service or repaired, the probability and
consequences of a SGTR are reduced. This RG uses safety factors on
loads for tube burst that are consistent with the requirements of
Section III of the ASME [American Society of Mechanical Engineers]
Code.
For axially oriented cracking located within the tubesheet, tube
burst is precluded due to the presence of the tubesheet. For
circumferentially oriented cracking, WCAP [Westinghouse Commercial
Atomic Power] --16506-P defines a length of degradation free
expanded tubing that provides the necessary resistance to tube
pullout due to the pressure induced forces (with applicable safety
factors applied). Application of the limited tubesheet inspection
depth criteria will preclude unacceptable primary-to-secondary
leakage during all plant conditions. The methodology for determining
leakage provides for large margins between calculated and actual
leakage values in the proposed limited tubesheet inspection depth
criteria.
Plugging of the SG tubes reduces the reactor coolant flow margin
for core cooling. Implementation of the 17 inch inspection length at
Turkey Point Units 3 and 4 will result in maintaining the margin of
flow that may have otherwise been reduced by tube plugging.
Based on the above, it is concluded that the proposed changes do
not result in any reduction of margin with respect to plant safety
as defined in the UFSAR or Bases of the plant Technical
Specifications.
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: Michael L. Marshall, Jr.
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: November 14, 2005.
Description of amendment request: The proposed amendment revised
the table of Primary Containment Isolation Instrumentation to eliminate
the trip generated by the main steamline radiation monitors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change deletes the Main Steamline Radiation Monitor
(MSLRM) trip function from TS [technical specification]. The MSLRM
is not an initiator of any accident previously evaluated. As a
result, the probability of any accident previously evaluated is not
significantly increased. The consequences of an accident previously
evaluated, specifically the Control Rod Drop Accident (CRDA), have
been evaluated consistent with the DAEC [Duane Arnold Energy Center]
licensing basis utilizing the Alternative Source Term (10 CFR
50.67). As demonstrated by the dose calculations, the consequences
of the accident are within the regulatory acceptance criterion. As a
result, the consequences of any accident previously evaluated are
not significantly increased. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
No new or different accidents result from utilizing the proposed
change. The changes do not involve a change in the methods governing
normal plant operation. The equipment proposed to be removed from
the plant, the MSLRM, is only credited in the CRDA analysis and no
other event in the safety analysis. The proposed changes are
consistent with the revised safety analysis assumptions for a CRDA
included in this application.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change deletes the requirement for the MSLRM
isolation function. Analyses performed consistent with the DAEC
licensing basis, demonstrate that the removal of this isolation will
not cause a significant reduction in the margin of safety, as the
resulting offsite dose consequences are being maintained within
regulatory limits.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. R.E. Helfrich, Florida Power & Light
Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: L. Raghavan.
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: December 22, 2005.
Description of amendment request: The proposed amendment revised
the reactor-pressure vessel material surveillance program described
within the Duane Arnold Energy Center (DAEC) Updated Final Safety
Analysis Report from a plant-specific program to the Boiling Water
Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance
Program (ISP).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change implements an integrated surveillance
program that has been evaluated by the NRC [Nuclear Regulatory
Commission] staff as meeting the requirements of paragraph III.C of
Appendix H to 10 CFR 50. Consequently, the proposed change does not
significantly increase the probability of any accident previously
evaluated. The proposed change provides the same assurance of RPV
[reactor pressure vessel] integrity. As a result, the consequences
of any accident previously evaluated are not significantly
increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change revises the DAEC licensing bases to reflect
participation in the BWRVIP ISP. The ISP was approved by the NRC
staff as an acceptable material surveillance program which complies
with 10 CFR 50, Appendix H. The proposed change maintains an
equivalent level of RPV material surveillance and does not introduce
any new accident initiators. The proposed change will not impact the
manner in which the plant is designed or operated. This change will
not affect the reactor pressure vessel, as no physical changes are
involved. The proposed change will not cause the reactor pressure
vessel or interfacing systems to be operated outside of any design
or testing limits. Furthermore, the proposed changes will not alter
any assumptions
[[Page 43534]]
previously made in evaluating the radiological consequences of any
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 amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change has been evaluated as providing an
acceptable alternative to the plant-specific RPV material
surveillance program that meets the requirements of the regulations
for RPV material surveillance. The material surveillance program
requirements contained in 10 CFR 50, Appendix H provide assurance
that adequate margins of safety exist for the reactor coolant system
against nonductile or rapidly propagating failures during normal
operation, anticipated operational occurrences, and system
hydrostatic tests.
The BWRVIP ISP has been approved by the NRC staff as an
acceptable material surveillance program which complies with I0 CFR
50, Appendix H. The ISP will provide the material surveillance data
which will ensure that the safety margins required by NRC
regulations are maintained for the DAEC reactor coolant system.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. R. E. Helfrich, Florida Power & Light
Company, P. O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: L. Raghavan.
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: April 28, 2006.
Description of amendment request: The proposed amendment modified
technical specifications (TSs) requirements for inoperable snubbers by
adding Limiting Condition for Operation (LCO) 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 28, 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. 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. 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 operation. 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: Mr. R.E. Helfrich, Florida Power & Light
Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: L. Raghavan.
Indiana Michigan Power Company, Docket No. 50-315, Donald C. Cook
Nuclear Plant, Unit 1, Berrien County, Michigan
Date of amendment request: April 10, 2006.
Description of amendment request: The proposed amendment revised
Surveillance Requirement 3.8.1.11 of the Donald C. Cook Technical
Specifications, raising the emergency diesel generator full load
rejection voltage test limit from 5000 volts to 5350 volts.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee provided a
no significant hazards determination analysis, which is reproduced
below:
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Response: No.
Probability of Occurrence of an Accident Previously Evaluated.
The proposed change is an increase in the Technical
Specification (TS) Surveillance Requirement (SR) limit on maximum
voltage following an emergency diesel generator (DG) full load
rejection. The DGs' safety function is solely mitigative and is not
needed unless there is a loss of offsite power. The DGs do not
affect any accident initiators or precursors of any accident
previously evaluated. The proposed increase in the TS SR limit does
not affect the DGs' interaction with any system whose failure or
malfunction can initiate an accident. Therefore, the probability of
occurrence of an accident previously evaluated is not significantly
increased.
Consequences of an Accident Previously Evaluated.
The DG safety function is to provide power to safety related
components needed to mitigate the consequences of an accident
following a loss of offsite power. The purpose of the TS SR voltage
limit is to assure DG damage protection following a full load
rejection. The technical analysis performed to support this proposed
amendment has demonstrated that the DGs can withstand voltages above
the new proposed limit without a loss of protection. The proposed
higher limit will continue to provide assurance that the DG is
protected, and the safety function of the DG will be unaffected by
the proposed change. Therefore, the consequences of an accident
previously evaluated will not be significantly increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
There are no new DG failure modes created and the DGs are not an
initiator of any new
[[Page 43535]]
or different kind of accident. The proposed increase in the TS SR
limit does not affect the interaction of the DGs 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.
The margins of safety applicable to the proposed change are
those associated with the ability of the DGs to perform their safety
function. The technical analysis performed to support this amendment
demonstrates that this ability will be unaffected. The increase in
the TS SR limit will not affect this ability. Therefore, the
proposed change does not involve a significant reduction in margin
of safety.
The NRC staff evaluated the licensee's analysis, and based on
this evaluation, the NRC staff proposes to determine that the
requested amendment does not involve a significant hazards
consideration.
Attorney for licensee: James M. Petro, Jr., Esquire, One Cook
Place, Bridgman, MI 49106.
NRC Branch Chief: L. Raghavan.
Nebraska Public Power District (NPPD), Docket No. 50-298, Cooper
Nuclear Station, Nemaha County, Nebraska
Date of amendment request: June 16, 2006.
Description of amendment request: The proposed amendment revised
Technical Specification (TS) 3.10.1, ``Inservice Leak and Hydrostatic
Testing Operation,'' to extend the scope to include provisions for
temperature increases above 212 [deg]F as a consequence of inservice
leak or hydrostatic testing, and as a consequence of control rod scram
time testing initiated in conjunction with the inservice leak test or
hydrostatic test, when initial test conditions are below 212 [deg]F.
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.
Current TS LCO [Limiting Condition for Operation] 3.10.1 allows
average RCS [reactor coolant system] temperature to exceed 212
[deg]F when required during the conduct of hydrostatic and inservice
leak tests without requiring entry into plant operating Mode 3, Hot
Shutdown. Extending this allowance to testing in which average RCS
temperature exceeds 212 [deg]F as a consequence of maintaining
pressure and to the performance of scram time testing that is
initiated in conjunction with the hydrostatic and inservice leak
tests will not impact any accident initiator. Thus, the proposed
change does not affect the probability of any accident.
The proposed changes do not involve any modification of
equipment used to mitigate accidents, and do not impact any system
used in the mitigation of design basis accidents. The proposed
changes do not involve modified operation of equipment or [a] system
used to mitigate accidents. Thus, the proposed changes do not affect
the consequences of an accident.
Based on the above, NPPD concludes that 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 TS revisions to TS LCO 3.10.1 do not involve
physical modification of the plant or a change in plant operation.
The proposed TS revisions do not revise or eliminate any existing
requirements, and do not impose any additional requirements. The
proposed changes do not alter assumptions made in the safety
analysis, and are consistent with the safety analysis assumptions
and current plant operating practice. Allowing the performance of
control rod scram time testing, while in plant operating Mode 4 with
average RCS temperature greater than 212 [deg]F, does not create the
possibility of a different kind of accident.
Based on the above NPPD[,] concludes that these proposed changes
do not create the possibility of a new or different kind of accident
from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not impact the design or operation of
the Reactor Protection System or the Emergency Core Cooling System.
Allowing completion of scram time testing that was initiated in
conjunction with inservice leak or hydrostatic testing prior to
reactor criticality and startup will eliminate the need for
unnecessary plant maneuvers to control reactor temperature and
pressure, thereby resulting in enhanced safe operation.
Based on the above, NPPD concludes that these proposed changes
do not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: David Terao.
Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile
Point Nuclear Station Unit No. 1, Oswego County, New York
Date of amendment request: January 18, 2006.
Description of amendment request: The proposed amendment deleted
the reference to the hydrogen monitors in Technical Specification (TS)
3.6.11, ``Accident Monitoring Instrumentation'' consistent with the
NRC-approved Industry/Technical Specification Task Force (TSTF)
Standard Technical Specification Change Traveler, TSTF-447,
``Elimination of Hydrogen Recombiners and Change to Hydrogen and Oxygen
Monitors.''
The NRC staff issued a notice of availability of ``Model
Application Concerning Technical Specification Improvement To Eliminate
Hydrogen Recombiner Requirement, and Relax the Hydrogen and Oxygen
Monitor Requirements for Light Water Reactors Using the Consolidated
Line Item Improvement Process (CLIIP)'', in the Federal Register on
September 25, 2003 (68 FR 55416). The notice included a model safety
evaluation (SE), a model no significant hazards consideration (NSHC)
determination, and a model application.
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 confirming the applicability of the model NSHC
determination to NMP-1 and incorporating it by reference in its
application. The model NSHC determination 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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen [and
[[Page 43536]]
oxygen] monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
RG [Regulatory Guide] 1.97 Category 1, is intended for key variables
that most directly indicate the accomplishment of a safety function
for design-basis accident events. The hydrogen [and oxygen] monitors
no longer meet the definition of Category 1 in RG 1.97. As part of
the rulemaking to revise 10 CFR 50.44 the Commission found that
Category 3, as defined in RG 1.97, is an appropriate categorization
for the hydrogen monitors because the monitors are required to
diagnose the course of beyond design-basis accidents. [Also, as part
of the rulemaking to revise 10 CFR 50.44, the Commission found that
Category 2, as defined in RG 1.97, is an appropriate categorization
for the oxygen monitors, because the monitors are required to verify
the status of the inert containment.]
The regulatory requirements for the hydrogen [and oxygen]
monitors can be relaxed without degrading the plant's, emergency
response. The emergency response, in this sense, refers to the
methodologies used in ascertaining the condition of the reactor
core, mitigating the consequences of an accident, assessing and
projecting offsite releases of radioactivity, and establishing
protective action recommendations to be communicated to offsite
authorities. Classification of the hydrogen monitors as Category 3,
[classification of the oxygen monitors as Category 2] and removal of
the hydrogen [and oxygen] monitors from TS will not prevent an
accident management strategy through the use of the SAMGs [severe
accident management guidelines], the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen [and oxygen] monitor
requirements, including removal of these requirements from TS, does
not involve a significant increase in the probability or the
consequences of any 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 elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen [and oxygen] monitor requirements,
including removal of these requirements from TS, will not result in
any failure mode not previously analyzed. The hydrogen recombiner
and hydrogen [and oxygen] monitor equipment was intended to mitigate
a design-basis hydrogen release. The hydrogen recombiner and
hydrogen [and oxygen] monitor equipment are not considered accident
precursors, nor does their existence or elimination have any adverse
impact on the pre-accident state of the reactor core or post
accident confinement of radionuclides within the containment
building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in [a] Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen [and oxygen] monitor requirements,
including removal of these requirements from TS, in light of
existing plant equipment, instrumentation, procedures, and programs
that provide effective mitigation of and recovery from reactor
accidents, results in a neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI [Three Mile Island],
Unit 2 accident can be adequately met without reliance on safety-
related hydrogen monitors.
[Category 2 oxygen monitors are adequate to verify the status of
an inerted containment.]
Therefore, this change does not involve a significant reduction
in [a] margin of safety. [The intent of the requirements established
as a result of the TMI, Unit 2 accident can be adequately met
without reliance on safety-related oxygen monitors.]
Removal of hydrogen [and oxygen] monitoring from TS will not
result in a significant reduction in their functionality,
reliability, and availability.
The NRC staff has reviewed the model NSHC determination and its
applicability to NMP-1. Based on this review, the NRC staff proposes
to determine that the amendment request involves no significant
hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Richard J. Laufer.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: June 7, 2006.
Description of amendment request: The amendment deleted Required
Action D.1.2 in Technical Specification (TS) 3.7.10, ``Control Room
Emergency Ventilation System (CREVS),'' and Required Action C.1.2 in TS
3.7.11, ``Control Room Air Conditioning System (CRACS).'' These
required actions are for the condition where the required actions and
completion time (CT) of TS 3.7.10 Condition A (one CREVS train
inoperable) and TS 3.7.11 Condition A (one CRACS train inoperable) are
not met in Modes 5 or 6, or during movement of irradiated fuel
assemblies. The deleted required actions, and associated CTs, are to
verify the operable CREVS (or CRACS) train is capable of being powered
by an emergency power source.
The amendment would also delete the phrase ``in MODES 1, 2, 3, or
4'' from Condition A (one emergency exhaust system (EES) train
inoperable) of TS 3.7.13, ``Emergency Exhaust System (EES),'' and
revise Condition D to state the following: ``Required Action and
associated Completion Time of Condition A not met during movement of
irradiated fuel assemblies in the fuel 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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Incorporation of a 7-day Completion Time for restoring an
inoperable EES train during shutdown conditions (i.e., during
movement of irradiated fuel assemblies in the fuel building) and the
deletion of Required Actions for verifying the availability of an
emergency power source when a CREVS/CRACS train is inoperable during
the same [shutdown] conditions, are operational provisions that have
no impact on the frequency of occurrence of the event for which the
EES, CREVS and CRACS are designed to mitigate, i.e., a fuel handling
accident (FHA) in the fuel building. These systems, (i.e., their
failure)[,] have no bearing on the occurrence of a fuel handling
accident as the systems themselves are not associated with any of
the potential initiating sequences, mechanisms or occurrences--such
as failure of a lifting device or crane [lifting a fuel assembly],
or an operator error--that could cause an FHA. Since these systems
are designed only to respond to an FHA as accident mitigators after
the accident has occurred, and they have no bearing on the
occurrence of such an event themselves, the proposed changes to the
CREVS, CRACS and EES Technical Specifications have no impact on the
probability of occurrence of an FHA. On this basis, the proposed
changes do not involve a significant increase in the probability of
an accident previously evaluated.
With regard to [the] consequences of previously evaluated
accidents (i.e., an FHA),
[[Page 43537]]
the proposed changes involve no design or physical changes to the
EES or any other equipment required for accident mitigation.
With respect to deleting the noted Required Actions (for
verifying that the operable CREVS/CRACS train is capable of being
powered from an emergency power source when on CREVS/CRACS train is
inoperable), such a change does not change the Limiting Condition
for Operation (LCO) requirement for both CREVS/CRACS trains to be
operable, nor to the LCO requirements of the TS requirements
pertaining to electrical power sources/support for shutdown
conditions. The change to the Required Actions would thus not be
expected to have a significant impact on the availability of the
CREVS and CRACS. That is, adequate availability may be still assumed
such that these systems would continue to be available to provide
their assumed [safety] function for limiting the dose consequences
of an FHA in accordance with the accident analysis currently
described in the FSAR [Callaway Final Safety Analysis Report].
With respect to the allowed outage time (Completion Time) for an
inoperable EES train, the consequences of a postulated accident are
not affected by equipment allowed outage times as long as adequate
equipment availability is maintained. The proposed EES allowed
outage time is based on the allowed outage time specified in the
Standard Technical Specifications (STS) for which it may be presumed
that the specified allowed outage time (Completion Time) is
acceptable and supports adequate EES availability. As noted in the
STS Bases, the 7-day Completion Time for restoring an inoperable EES
train takes into account the availability of the other train [(i.e.,
the other train is operable)]. Since the STS-supported Completion
Time supports adequate EES availability, it may be assumed that the
EES function would be available for mitigation of an FHA, thus
limiting offsite dose to within the currently calculated [dose
consequence] values based on the current accident analysis [in the
FSAR]. On this basis, the consequences of applicable, [previously]
analyzed accidents (i.e., the FHA) are not increased by the proposed
change.
Based on the above, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not create any new failure modes for any
system or component, nor do they adversely affect plant operation.
No hardware or design changes are involved. Thus, no new equipment
will be added and no new limiting single failures must be
postulated. The plant will continue to be operated within the
envelope of the existing safety analysis [in the FSAR].
Therefore, the proposed changes do not create [the possibility
of] a new or different kind of accident [from any accident]
previously evaluated.
3. Do the proposed change[s] involve a significant reduction in
a margin of safety?
Response: No.
The calculated radiological dose consequences per the applicable
accident analyses remain bounding since they are not impacted by the
proposed changes. The margins [of safety] to the limits of 10 CFR
100 [Title 10 of the Code of Federal Regulations Part 100] and GDC
[General Design Criterion] 19 [of Appendix A to 10 CFR Part 50] are
thus unchanged by the proposed changes.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Pillsbury Winthrop Shaw
Pittman LLP, 2300 N Street, NW., Washington, DC 20037.
NRC Branch Chief: David Terao.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: May 22, 2006.
Description of amendment request: The proposed amendment revised
Technical Specification (TS) 1.1, ``Definitions,'' TS 3.4.13, ``RCS
Operational LEAKAGE,'' TS 5.5.8, ``Steam Generator (SG) Program,'' and
TS 5.6.7, ``Steam Generator Tube Inspection Report,'' and adds TS
3.4.20, ``Steam Generator (SG) Tube Integrity.'' 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 licensee has evaluated whether or not a
significant hazards consideration is involved with the proposed changes
by focusing on the three standards set forth in 10 CFR 50.92,
``Issuance of Amendment,'' as discussed below:
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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change requires a SG Program that includes
performance criteria that will provide reasonable assurance that the
SG tubing will retain integrity over the full range of operating
conditions (including startup, operation in the power range, hot
standby, cooldown and all anticipated transients included in the
design specification). The SG performance criteria are based on tube
structural integrity, accident induced leakage, and operational
leakage.
A SG tube rupture (TR) event is one of the design basis
accidents that are analyzed as part of a plant's licensing basis. In
the analysis of a SGTR event, a bounding primary to secondary
leakage rate equal to the operational leakage rate limits in the
licensing basis plus the leakage rate associated with a double-ended
rupture of a single tube is assumed.
For other design basis accidents such as main steam line break
(MSLB), rod ejection, and reactor coolant pump locked rotor the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). These analyses typically assume that
primary to secondary leakage for all SGs is 1 gallon per minute or
increases to 1 gallon per minute as a result of accident induced
stresses. The accident induced leakage criterion introduced by the
proposed changes accounts for tubes that may leak during design
basis accidents. The accident induced leakage criterion limits this
leakage to no more than the value assumed in the accident analysis.
The SG performance criteria proposed change to the 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 TS 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 1 gallon per minute
with no more than 500 gallons per day in any one SG, and that the
reactor coolant activity levels of DOSE EQUIVALENT 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.
[[Page 43538]]
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.
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 TS. 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.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
SG 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 has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Branch Chief: Evangelos C. Marinos.
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].
Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power
Station, Kewaunee County, Wisconsin
Date of application for amendment: February 6, 2006, as
supplemented by letter dated May 5, 2006.
Brief description of amendment: The proposed amendment added a
license condition to extend certain Technical Specification (TS)
surveillance intervals on a one-time basis to account for the effects
of an extended forced outage in the spring of 2005.
Date of issuance: July 12, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 187.
Facility Operating License No. DPR-43: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in the Federal Register: March 14, 2006 (71
FR 13172).
The supplemental letter 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.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 12, 2006.
No significant hazards consideration comments received: No.
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 application of amendments: June 15, 2005.
Brief description of amendments: The amendments revised the
Technical Specifications to eliminate the out of date requirements
associated with the completion of the Keowee Refurbishment
modifications on both Keowee Hydro Units (KHUs).
Date of Issuance: July 11, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 353, 355, and 354.
Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Licenses and the Technical Specifications.
Date of initial notice in the Federal Register: May 9, 2006 (71 FR
26998).
[[Page 43539]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 11, 2006.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: July 5, 2005, as supplemented by
letter dated March 22, 2006.
Brief description of amendment: The amendment modified the existing
Technical Specification 3.3.1.3, ``Oscillation Power Range Monitor
(OPRM) Instrumentation,'' Surveillance Requirement 3.3.1.3.5.
Specifically, the thermal power level at which the OPRMs are ``not
bypassed'' (enabled to perform their design function) will be change
from > 28.6-percent rated thermal power to >= 23.8-percent rated
thermal power.
Date of issuance: June 30, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 138.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications and License.
Date of initial notice in the Federal Register: August 16, 2005 (70
FR 48206).
The March 22, 2006 supplement, contained clarifying information and
did not change the NRC staff's initial proposed finding of no
significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 30, 2006.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: March 7, 2006.
Brief description of amendments: The amendments revised Section
5.5.2, ``Leakage Monitoring Program,'' of the units'' Technical
Specifications, adding the Liquid Waste Disposal System, Waste Gas
System, and Post-Accident Containment Hydrogen Monitoring System to the
list of systems. The listing of these systems was inadvertently omitted
from Section 5.5.2.
Date of issuance: July 5, 2006.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 294 and 297.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revise the Technical Specifications and Licenses.
Date of initial notice in the Federal Register: April 11, 2006 (71
FR 18374).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 5, 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: June 29, 2005, as supplemented
by letter dated April 25, 2006.
Brief description of amendment: The amendment revised Technical
Specifications Table 3.3.8.1-1, ``Loss of Power Instrumentation,''
changing the allowable values for the 4.16-kV essential bus degraded
voltage from a range of 3897-3933 volts to a range of 3913-3927 volts.
Date of issuance: July 3, 2006.
Effective date: As of the date of issuance and shall be implemented
concurrently with implementation of the Improved Technical
Specifications (Amendment No. 146, dated June 5, 2006).
Amendment No: 147.
Facility Operating License No. DPR-22: Amendment revised the
Facility Operating License and Technical Specifications.
The supplemental letter 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 the Federal Register: November 23, 2005
(70 FR 70889).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 3, 2006.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of application for amendment: February 16, 2006.
Brief description of amendment: The amendment revised the Technical
Specifications to make the existing SG tube surveillance program
consistent with the Commission's approved Technical Specification Task
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
449, ``Steam Generator Tube Integrity,'' Revision 4.
Date of issuance: July 6, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 223.
Facility Operating License No. DPR-20: Amendment revised the
Technical Specifications and License.
Date of initial notice in the Federal Register: May 23, 2006 (71 FR
29679).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 6, 2006.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: November 11, 2005, supplemented
by letter dated March 23, 2006.
Brief description of amendments: The amendments revise PINGP's
Technical Specification (TS) 3.6.5, ``Containment Spray and Cooling
Systems,'' to incorporate changes to an existing condition and two
surveillance requirements, and also to add a new condition that will
allow continued plant operation with TS limitations when two
containment cooling system fan coil units, one in each train, are
inoperable.
Date of issuance: June 29, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 173 and 163.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in the Federal Register: February 28, 2006
(71 FR 10074).
The supplemental letter 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.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 29, 2006.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of application for amendment: January 19, 2006.
Brief description of amendment: The amendment revises the Humboldt
Bay Unit 3 Technical Specifications to correct an editorial error and
to allow leaving the Unit 3 control room temporarily unmanned during
[[Page 43540]]
emergency conditions requiring personnel to evacuate occupied buildings
for their safety.
Date of issuance: July 10, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 38.
Facility Operating License No. DPR-7: This amendment revised the
Technical Specifications and License.
Date of initial notice in the Federal Register: February 28, 2006
(71 FR 10077).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 10, 2006.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket No. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne County,
Pennsylvania
Date of application for amendments: February 1, 2006, as
supplemented on June 27, 2006.
Brief description of amendments: The amendments revise the
Technical Specification (TS) requirements for inoperable snubbers by
adding limiting condition for operation 3.0.8 for SSES 1 and 2. This
change is based on the TS Task Force (TSTF) change traveler TSTF-372,
Revision 4. A notice of availability for this TS improvement using the
consolidated line item improvement process was published in the Federal
Register on November 24, 2004, and May 4, 2005.
Date of issuance: July 7, 2006.
Effective date: As of the date of issuance and to be implemented
within 60 days.
Amendment Nos.: 236 and 213.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the Technical Specifications and License.
Date of initial notice in the Federal Register: April 25, 2006 (71
FR 23959).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 7, 2006.
The supplement dated June 27, 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.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: October 6, 2005, as supplemented April
17, 2006.
Brief Description of amendments: The amendments revised Technical
Specification (TS) Section 5.6.5, ``Core Operating Limits Report
(COLR),'' to reflect the addition of the methodology in WCAP-16009-P-A,
``Realistic Large Break LOCA [Loss-Of-Coolant Accident] Evaluation
Methodology Using the Automated Statistical Treatment of Uncertainty
Method (ASTRUM),'' for and provide a new large break LOCA analyses for
Farley Units 1 and 2.
Date of issuance: July 11, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 174/167.
Renewed Facility Operating License Nos. NPF-2 and NPF-8: Amendments
revise the Technical Specifications and Licenses.
Date of initial notice in the Federal Register: November 8, 2005
(70 FR 67751). The supplemental letter provided clarifying information
that was within the scope of the initial notice and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 11, 2006.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Unit Nos. 1 and 2, Houston County,
Alabama
Date of amendments request: February 17, 2006.
Brief Description of amendments: The amendments revised the
Technical Specifications (TSs) adding 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, if risk
is assessed and managed consistent with the program in place for
complying with the requirements of 10 CFR 50.65(a)(4).
Date of issuance: June 29, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 173/166.
Renewed Facility Operating License Nos. NPF-2 and NPF-8: Amendments
revised the Licenses and the Technical Specifications.
Date of initial notice in the Federal Register: April 25, 2006 (71
FR 23960).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 29, 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: December 16, 2005.
Brief description of amendments: The amendments revised the
Technical Specifications ACTIONS NOTE for TS 3.7.5, ``Auxiliary
Feedwater (AFW) System,'' based on Industry/Technical Specification
Task Force (TSTF) Standard Technical Specification Change Traveler
TSTF-359, Revision 9, ``Increased Flexibility in Mode Restraints.''
Date of issuance: July 14, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 142 and 122.
Facility Operating License Nos. NPF 68 and NPF-81: Amendments
revised the Licenses and the Technical Specifications.
Date of initial notice in the Federal Register: February 14, 2006
(71 FR 7813).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 14, 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 26, 2005, as supplemented by
letter dated March 9, 2006.
Brief description of amendment: The amendment revised TS 3.7.2,
``Main Steam Isolation Valves (MSIVs),'' by adding the MSIV actuator
trains to (1) the limiting condition for operation (LCO) and (2) the
conditions, required actions, and completion times for the LCO. The
existing conditions and required actions in TS 3.7.2 are renumbered to
account for the new conditions and required actions.
Date of issuance: June 16, 2006.
Effective date: 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-30: The amendment revised the
Technical Specifications and License.
[[Page 43541]]
Date of initial notice in the Federal Register: June 21, 2005 (70
FR 35740).
The supplemental letter dated March 9, 2006, provided additional
clarifying information, 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 published in the
Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 16, 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 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
[[Page 43542]]
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
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\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly 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.
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Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by E-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer or the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
Exelon Generation Company, LLC, Docket No. 50-353, Limerick Generating
Station (LGS), Unit 2, Montgomery County, Pennsylvania
Date of amendment request: June 9, 2006, as supplemented June 16,
and June 23, 2006.
Description of amendment request: The one-time amendment revises
Technical Specification (TS) Limiting Condition for Operation 3.6.1.7
concerning drywell average air temperature. Specifically, the proposed
change would add a footnote to the TS limit for drywell average air
temperature of 145 degrees Fahrenheit ([deg]F) to allow continued
operation of LGS, Unit 2, with drywell average air temperature no
greater than 148 [deg]F for the remainder of the current operating
cycle (Cycle 9), which is currently scheduled to end in March 2007, or
until the next shutdown of sufficient duration to allow for unit cooler
fan repairs, whichever comes first.
Date of issuance: July 7, 2006.
Effective date: As of date of issuance, to be implemented within 14
days.
Amendment No.: 145.
Facility Operating License No. NPF-85: The amendment revises the
Technical Specifications and License.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. June 20, 2006 (71 FR 35453). The notice
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. No comments have been received. The notice also
provided an opportunity to request a hearing by July 5, 2006, but
indicated that if the Commission makes a final NSHC determination, any
such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a safety evaluation dated July 7, 2006.
The supplements dated June 16 and June 23, 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.
Attorney for licensee: Mr. Brad Fewell, Assistant General Counsel,
Exelon Generation Company, LLC, 200 Exelon Way, Kennett Square, PA
19348.
NRC Branch Chief: Darrell J. Roberts.
Dated at Rockville, Maryland, this 25th day of July, 2006.
[[Page 43543]]
For the Nuclear Regulatory Commission.
Cornelius F. Holden,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 06-6597 Filed 7-31-06; 8:45 am]
BILLING CODE 7590-01-P