[Federal Register Volume 71, Number 224 (Tuesday, November 21, 2006)]
[Notices]
[Pages 67391-67403]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-19434]
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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 October 27, 2006, to November 8, 2006. The
last biweekly notice was published on November 7, 2006 (71 FR 65139).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or
[[Page 67392]]
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 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].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station (Oyster Creek), Ocean County, New Jersey
Date of amendment request: September 28, 2006.
Description of amendment request: The amendment would revise the
Oyster Creek Technical Specifications definition of Channel
Calibration, Channel Check, and Channel Functional Test in accordance
with the NUREG-1433, Revision 3, ``Standard Technical Specifications,
General Electric Plants--BWR [boiling water reactor]/4.''
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. Will operation of the facility in accordance with the
proposed amendment involve a significant increase in the probability
or consequences of an accident previously evaluated?
Response: No.
The definitions of Channel Check, Channel Calibration[,] and
Channel Functional Test specified in Technical Specifications (TS)
provide basic information regarding what the test involves, the
components involved in the test, and general information regarding
how the test is to be performed. Instrument channel checking,
calibrating, and testing are not initiators of any accident
previously evaluated. Furthermore, the proposed changes will not
affect the ability of the channel being checked, calibrated[,] or
tested to respond as assumed in any accident previously evaluated.
Therefore, these revised definitions result in no increase in the
probability of an accident previously evaluated.
The proposed revisions of these definitions, corresponding
administrative changes (capitalization of definitions), and the
proposed alternate testing and calibrating methodology using
sequential, overlapping testing, and/or actual channel input signals
and/or in place qualitative assessments of resistance temperature
detectors (RTD's) and thermocouples (TC's) involve no changes to
plant design, equipment, or operation related to mitigation of
accidents. The qualitative evaluation of sensor behavior for non-
adjustable sensors will provide an accurate indication of sensor
operation and will
[[Page 67393]]
assure that [the evaluated] portion of the channel is operating
properly, ensuring that the consequences of an accident will remain
as previously evaluated. Therefore, these revised definitions result
in no increase in the consequences of an accident previously
identified.
Based on the above, AmerGen concludes that the proposed changes
do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Will operation of the facility in accordance of the proposed
amendment create the possibility of a new or different kind of
accident from any accident previously evaluated?
Response: No.
The proposed revisions of the instrument surveillance
definitions, corresponding administrative changes (capitalization of
definitions), and the proposed alternate testing and calibrating
methodology using sequential, overlapping testing, and/or actual
channel input signals and/or in place qualitative assessments of
RTD's and TC's do not involve a physical alteration of the plant or
a change in the methods governing normal plant operation. No new or
different type[s] of equipment will be installed. The proposed
changes also do not adversely affect the operation or operability of
existing plant equipment. The proposed revisions will allow a change
in testing and calibrating methodology. Allowing an alternate
testing and calibrating methodology will not change how the plant is
operated. Each instrument channel will be tested one sub channel at
a time, as is currently performed, and will not create the
possibility of a new or different kind of accident.
Based on the above discussion, AmerGen concludes that the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Will operation of the facility in accordance with the
proposed amendment involve a significant reduction in a margin of
safety?
Response: No.
The affected definitions involve checking, calibrating[,] and
testing of instrumentation used in the mitigation of accidents to
ensure that the instrumentation will perform as assumed in safety
analyses. The proposed revisions of these definitions, corresponding
administrative changes (capitalization of definitions), and the
proposed alternate testing and calibrating methodology using
sequential, overlapping testing, and/or actual channel input signals
and/or in place qualitative assessments of RTD's and TC's does not
alter the ability of the instrument channel to respond as designed
or assumed in the safety analyses. As a result[,] the ability of the
plant to respond to[,] and mitigate[,] accidents is unchanged by the
revised definitions. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of amendment request: June 16, 2006, as supplemented by letter
dated September 14, 2006.
Description of amendment request: The proposed amendment would
revise the Byron Station Updated Final Safety Analysis Report (UFSAR)
to incorporate changes concerning the requirements for physical
protection from tornado-generated missiles (TGM) for safety-related and
non-safety related systems and components.
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 probability of occurrence of the design basis tornado
remains the same as originally established in the Byron Station
Updated Final Safety Analysis Report (UFSAR). The request involves
the use of a probability-based assessment of the need for physical
tornado missile protection of specific existing features at Byron
Station.
The request is to utilize an NRC approved methodology (i.e., the
Electric Power Research Institute (EPRI) Topical Report ``Tornado
Missile Risk Evaluation Methodology'') to conclude that the
acceptance criteria of NUREG-0800, ``Standard Review Plan,'' (SRP)
Section 2.2.3, ``Evaluation of Potential Accidents,'' Revision 2,
July 1981, has been met for Byron Station and that tornado missile
damage of selected components at Byron Station need not be
considered as a credible event.
Per Item 2 in Section III of SRP 3.5.1.4, probability methods
can be used to accept tornado missile effects provided damage to all
important structures, systems and components, as discussed in
Regulatory Guide 1.117 are considered. Per Section II of the SRP,
the acceptance criterion of SRP 2.2.3 is applicable. Section II of
SRP 2.2.3 states that the expected rate of occurrence of potential
exposure in excess of 10 CFR Part 100, ``Reactor Site Criteria,''
guidelines of approximately 1.0E-06 per reactor year is acceptable,
if when combined with reasonable qualitative arguments, that the
realistic probability can be shown to be smaller.
[The licensee in its September 14, 2006, letter stated the
following in regards to the consequences of an accident previously
evaluated:
The acceptance criteria for the TORMIS analysis has been
established as 1.0 E-06 per year cumulative probability of a TGM
striking/damaging an unprotected essential SSC [system, structure or
component] required for safe shutdown in the event of a tornado,
which is the same value found to be acceptable by the NRC based on
the accepted rates of occurrence of potential exposures in excess of
10 CFR 100 guidelines. This criteria in combination with
conservative qualitative assumptions show that the realistic
probability of a potential exposure in excess of the 10 CFR Part 100
guidelines is lower than 1.0 E-06 per year. The conservative
qualitative assumptions are the same as previously found to be
acceptable by the NRC as described below:
It is assumed that an essential SSC being struck/damaged by a
tornado missile will result in damage sufficient to preclude it from
performing its safety function.
It is assumed that the damage to the essential SSC results in
damage to fuel sufficient to result in conservatively calculated
radiological release values in excess of 10 CFR 100 guidelines.
There are no missiles that can directly impact irradiated fuel,
even the spent fuel stored in the Spent Fuel Pool.]
The proposed change is not considered to constitute a
significant increase in the probability or occurrence or the
consequences of an accident due to the extremely low probability of
damage due to tornado-generated missiles and therefore an extremely
low probability of a radiological release. Therefore, the proposed
changes do not involve a significant increase in the probability or
consequences of previously evaluated accidents.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This change involves the use of an alternative methodology to
assess the need for tornado missile protection on selected Byron
Station components. The use of this methodology and the changes to
the Byron Station UFSAR will be limited to design basis tornado
applications and do not contribute to the possibility of a new or
different kind of accident from those previously analyzed.
No new or different system interactions are created and no new
processes are introduced. The proposed change does not introduce any
new failure mechanisms, malfunctions, or accident initiators not
already considered in the design and licensing bases. Based on this
evaluation, the proposed change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
[[Page 67394]]
The changes, allowing for no additional physical protection for
tornado-generated missiles for certain Byron Sation components, is
based on successfully meeting the acceptance criteria of NUREG-0800,
``Standard Review Plan,'' (SRP) Section 2.2.3, ``Evaluation of
Potential Accidents,'' Revision 2, July 1981. Because of the
extremely low probability of damage to these components from
tornado-generated missiles, the change is not considered to
constitute a significant decrease 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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Daniel S. Collins.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1, Lake County, Ohio
Date of amendment request: October 13, 2006.
Description of amendment request: The proposed amendment would
eliminate License Condition 2.F, which requires reporting violations of
Operating License Section 2.C, and eliminates Technical Specification
5.6.6, which contains a reporting condition similar to Operating
License Section 2.C.(6).
The availability of this operating license improvement was
announced in the Federal Register on November 4, 2005 (70 FR 67202), as
part of the consolidated line item improvement process (CLIIP). The NRC
staff issued a notice of opportunity for comment in the Federal
Register on August 29, 2005 (70 FR 51098), on possible amendments
concerning this CLIIP, including a model safety evaluation and a model
no significant hazards consideration (NSHC) determination. The NRC
staff subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on November 4, 2005 (70 FR 67202). In its application dated October 13,
2006, the licensee affirmed the applicability of the following
determination.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change involves the deletion of a reporting
requirement. The change does not affect plant equipment or operating
practices and therefore does not significantly increase the
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No.
The proposed change is administrative in that it deletes a
reporting requirement. The change does not add new plant equipment,
change existing plant equipment, or affect the operating practices
of the facility. Therefore, the change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change deletes a reporting requirement. The change
does not affect plant equipment or operating practices and therefore
does not involve a significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Daniel S. Collins.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: October 5, 2006.
Description of amendment request: The proposed amendment to the
Improved Technical Specification will revise the defined pool burnup-
enrichment requirements, storage configuration for fresh fuel and low
burnup/high enriched fuel, the definition of a peripheral assembly, and
will include minor editorial changes.
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) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The LAR proposes to revise the fresh fuel loading configuration.
PEF [Progress Energy Florida, Inc.] has reanalyzed the criticality
of the revised storage configuration for fresh fuel checkerboarded
with spent fuel in Pool A, and surrounded by empty water cells in
Pool B. Similarly, storage of spent fuel in peripheral storage
locations, given the new definition, was also reanalyzed. The
revised fuel storage configuration does not affect any structure,
system, component or process related to the operation of Crystal
River Unit 3 (CR-3). As a result, the proposed LAR will not change
the probability or consequences of any accidents previously
evaluated that are related to operation of the plant. Thus, only
those accidents that are related to movement and storage of fuel
assemblies could be potentially affected by the proposed LAR.
Fuel Handling Accidents (FHAs) are analyzed in Section 14.2.2.3
of the CR-3 Final Safety Analysis Report (FSAR). These include a FHA
inside the Reactor Building (RB) and outside the RB. This LAR
involves storage of fuel assemblies, an activity conducted outside
the RB only. Therefore, only the FHA outside the RB event needs to
be considered.
The FHA outside the RB event is described as the dropping of a
fuel assembly into the spent fuel storage pool that results in
damage to a fuel assembly and the release of the gaseous fission
products. The current FHA assumes all 208 fuel pins in the dropped
assembly are damaged and the gas gap activity released. The results
of that analysis demonstrate that the applicable dose acceptance
criteria, 10 CFR 50.67 and Regulatory Guide 1.183, ``Alternative
Radiological Source Terms for Evaluating Design Basis Accidents at
Nuclear Power Reactors,'' are satisfied. Thus, the consequences of a
FHA are not increased by the allowed change in the fresh fuel
configuration. The fresh fuel storage configurations permit more
effective use of already existing storage locations. They do not
change the frequency or method for handling fuel assemblies. Fuel
handling equipment is unaffected. As such, the probability of a FHA
has not increased. Since only one fuel assembly is handled at a
time, the consequences of a FHA have not increased.
The current limiting heat load for the spent fuel pool is from
the combined impact of stored spent fuel and a full core off-load.
These changes do not increase spent fuel storage capacity over that
for which the racks are currently analyzed and it does not increase
the amount of heat ejected from an off-loaded core. Consequently,
current analyses for spent fuel pool cooling remain valid. The
configuration change allows fresh fuel to be checkerboarded with
spent fuel. Since these changes do not increase the storage capacity
over that already analyzed for the racks, filling the empty water
cells in the checkerboard pattern with spent fuel will not increase
the heat load over that already analyzed. The Pool B allowance to
surround a higher enriched/lower burnup fuel assembly in Pool B with
empty water cells or changing the definition of a periphery rack
cell does not increase the number of spent fuel assembly rack
locations over that previously analyzed. Therefore, there is no
increase in the pool heat load over that already analyzed.
A change in storage configurations in storage Pools A and B does
not increase the probability of a full core off-load or the
frequency of establishing maximum heat load conditions.
The FSAR specifies the normal upper limit of the fuel pool
cooling system as 160 [deg]F. Administrative controls are
implemented to
[[Page 67395]]
control when fuel may be moved from the reactor to the fuel pool to
prevent reaching this limit.
Because neither the probability nor the consequences of a FHA
are increased, and because there is not additional heat input to the
spent fuel pools, it is concluded that the LAR does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated?
Onsite storage of spent fuel assemblies in the spent fuel pools
is a normal activity for which CR-3 has been designed and licensed.
As part of assuring that this normal activity can be performed
without endangering public health and safety, the ability of CR-3 to
safely accommodate different possible accidents in the spent fuel
pools, such as dropping a fuel assembly or the misloading of a fuel
assembly, have been analyzed. The revised fuel storage
configurations proposed by the LAR does not change the methods of
fuel movement or fuel storage. No structural or mechanical change to
racks or fuel handling equipment is being proposed. The proposed
revisions allow for more effective use of existing, unmodified rack
locations when fresh or highly enriched, low burnup fuel is stored
in the pool. The proposed revisions are a modification to the
criticality analysis only, and therefore the proposed LAR does not
create any new or different kind of accident from those previously
evaluated.
(3) Involve a significant reduction in a margin of safety?
The CR-3 Improved Technical Specification (ITS) ensures the
effective neutron multiplication factor, Keff, of the spent fuel
storage racks is maintained less than or equal to 0.95 when fully
loaded and flooded with unborated water. The revisions proposed by
the LAR likewise ensure Keff is maintained less than this
requirement.
Analyses for the proposed fuel storage configurations have shown
that sufficient margin exists for fuel enriched to the maximum
allowed by the CR-3 license, and for all fuel that is or has been in
use at CR-3. Maintaining this margin is assured by remaining within
the limits on initial enrichment and fuel burnup that are specified
in the CR-3 ITS and, in the case of highly enriched, low burnup fuel
in Pool B, by water hole spacing. The LAR proposes allowing fresh
fuel to be checkerboarded with Category B type fuel in Pool A rather
than with empty water cells. It also allows fresh fuel with high
initial enrichment which does not meet current burnup requirements
to be placed in Pool B if surrounded by eight empty water cells. It
also proposes to change the definition of a periphery rack location
for storing Category BP type fuel. Analyses show that the new
proposed limits ensure that Keff remains less than 0.95. Attachment
E [not included in this FR notice] provides an analysis summary.
The current CR-3 licensing basis allows the use of
administrative controls, e.g., curves of initial fuel assembly
enrichment versus burnup, as a means of preventing criticality in
the spent fuel pools. The use of these curves would be continued
under this proposed amendment. The changes to these curves proposed
by this LAR consist of revising the values of burnup and adding
notes to restrict loading of certain fuel assemblies to specific
configurations. These types of curves and administrative controls
have been included in the CR-3 operating license and their use
implemented by site procedures for many operating cycles. From this
previous use, CR-3 personnel are familiar with the practice of using
administrative controls, such as curves of fuel assembly enrichment
versus burnup, to prevent criticality events when placing fuel
assemblies in the spent fuel pool.
Misloaded and mislocated fuel assemblies were analyzed. The
analysis demonstrated that misloading of a fresh fuel assembly,
assuming no soluble poison (boron) in the water does result in
exceeding the criticality margin regulatory limit of Keff = 0.95.
The analysis further shows that a concentration of 165 ppm boron in
the Pool A and a concentration of 46 ppm boron in Pool B is
sufficient to ensure Keff < 0.95. LCO 3.7.14 currently requires a
minimum boron concentration of 1925 ppm in the spent fuel pools
until fuel is verified as having been loaded in accordance with the
enrichment and burnup requirements of LCO 3.7.15. The soluble boron
assumed in the analysis for this proposed change is substantially
less than the 1925 ppm required by the existing license. Therefore,
existing license requirements for soluble boron remain conservative.
The NRC staff has reviewed the analysis provided for Florida Power
Corporation and, based on this review, it appears that the three
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief (Acting): L. Raghavan.
FPL Energy Duane Arnold, LLC, Docket No. 50-331, Duane Arnold Energy
Center (DAEC), Linn County, Iowa
Date of amendment request: July 17, 2006.
Description of amendment request: The proposed amendment would
revise the Limiting Condition for Operation (LCO) 3.6.3.1 to eliminate
the requirement for the Containment Atmospheric Dilution (CAD) system,
allowing its removal from the DAEC. LCO 3.6.3.2 would also be revised
to allow an additional 48 hours on plant start-up or shutdown sequences
for the primary containment to be de-inerted.
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 Containment Atmosphere Dilution (CAD) system and primary
containment oxygen concentration are not initiators to any accident
previously evaluated in the DAEC Updated Final Safety Analysis
Report (UFSAR). The CAD system and containment oxygen concentration
were previously relied upon to mitigate the consequences of a design
basis accident (DBA) combustible gas mixture. However, the revised
10 CFR 50.44 (68 FR 54123) no longer defines a DBA hydrogen release
(i.e., combustible gas mixture) and the Commission has subsequently
found that the DBA loss of coolant accident (LOCA) hydrogen release
is not risk significant. In addition, hydrogen control systems, such
as CAD, have been determined to be ineffective at mitigating
hydrogen releases from the more risk significant beyond design basis
accidents that could threaten containment integrity. Therefore,
elimination of the CAD system will not significantly increase the
consequences of any accident previously evaluated. The consequences
of an accident while relying on the revised Required Actions for
primary containment oxygen concentration are no different than the
consequences of the same accidents under the current Required
Actions. 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 physical alteration of the
plant, except for the elimination of the CAD system (i.e., no new or
different type of equipment will be installed) or a change in the
methods governing normal plant operation. The CAD system is not
considered an accident precursor, nor does its 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 from any DBA. In addition, the changes do
not impose any new or different requirements. The changes to the
Technical Specifications for oxygen concentration do not alter
assumptions made in the safety analysis, but reflect changes to the
safety analysis requirements allowed under the revised 10 CFR 50.44.
Specifically that an inerted containment is no[t] required to
mitigate any DBA, but has been found to be helpful in mitigating
certain beyond design basis events (i.e., severe accidents) that
could generate combustible levels of hydrogen.
[[Page 67396]]
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 installation of combustible gas control systems, such as
CAD, required by the original Sec. 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. (68 FR 54123). The proposed changes to CAD and primary
containment oxygen concentration reflect this new regulatory
position and, in light of the remaining plant equipment,
instrumentation, procedures, and programs that provide effective
mitigation of and recovery from reactor accidents, including
postulated beyond design basis events, does not result in a
significant reduction in a margin of safety.
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.
Indiana Michigan Power Company (I&M), Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit 2, Berrien County, Michigan
Date of amendment request: September 15, 2006.
Description of amendment request: The proposed amendment would
replace the current control system and it will increase the nominal
control fluid oil operating pressure from 114 pounds per square inch
gauge (psig) to 1600 psig. The control fluid oil pressure provides an
input to the reactor protection system via three pressure switches
connected to the control fluid header. Due to the change in the
operating pressure, I&M is proposing a revision to the allowable low
fluid oil pressure value from greater than or equal to 57 psig to
greater than or equal to 750 psig.
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 reflects a design change to the turbine
control system that increases the control oil pressure,
necessitating a change to the value at which a low fluid oil
pressure initiates a reactor trip. The turbine control oil pressure
is an input to the reactor trip instrumentation, and the reactor
trip is a response to an event that trips the turbine. A change in
the nominal control oil pressure does not introduce any mechanisms
that would increase the probability of an accident previously
analyzed. The reactor trip on turbine trip function is an
anticipatory trip, and the safety analysis does not credit this trip
for protecting the reactor core. Thus, the consequences of
previously analyzed accidents are not impacted.
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 control fluid oil pressure decreases in response to a
turbine trip. The value at which the low control fluid oil initiates
a reactor trip is not an accident initiator. The change in the value
reflects the higher pressure of the turbine control system that will
be installed during the Unit 2 Cycle 17 refueling outage.
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 change involves a parameter that initiates an anticipatory
reactor trip following a turbine trip. The safety analyses do not
credit this anticipatory trip for reactor core protection.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The Nuclear Regulatory Commission (NRC) staff has reviewed the
licensee's analysis and, based on this review, it appears that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the amendment requests involve no
significant hazards consideration.
Attorney for licensee: James M. Petro, Jr., Esquire, One Cook
Place, Bridgman, MI 49106.
NRC Acting Branch Chief: Martin C. Murphy.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Unit 1 and 2, Berrien County, Michigan
Date of amendment request: September 15, 2006.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TS) to change Required Action
Notes in TS 3.3.1, ``Reactor Trip System Instrumentation,'' and TS
3.3.2, ``Engineered Safety Features Actuation System Instrumentation,''
to reflect installed bypass test capability, as well as correct one
administrative error in TS 3.3.1 Condition Q. The proposed changes to
the Required Action Notes are consistent with wording in Standard
Technical Specifications (NUREG-1431, Revision 3) for plants with
installed bypass test capability.
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 of occurrence or consequences of an accident
previously evaluated?
Response: No.
The proposed change reflects NUREG-1431, Revision 3, ``Standard
Technical Specifications, Westinghouse Plants,'' (STS) wording for
plants with installed bypass test capability and aligns Technical
Specification (TS) Condition entry requirements with other portions
of the TS. The proposed changes do not modify how the reactor trip
system (RTS) and engineered safety features actuation systems
(ESFAS) functions respond to an accident condition. The proposed
changes to the TS Required Action Notes prevent unnecessary TS
Action entry during performance of surveillance testing. The
probability of accidents previously evaluated remains unchanged
since the proposed change does not affect any accident initiators.
The consequences of accidents previously evaluated are unaffected by
this change because no change to any accident mitigation scenario
has resulted and there are no additional challenges to fission
product barrier integrity.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No changes are being made to the plant that would introduce any
new accident causal mechanisms. The proposed change to
[[Page 67397]]
the Required Action Notes and Condition entry requirements does not
adversely affect previously identified accident initiators and does
not create any new accident initiators. The change does not affect
how the RTS and ESFAS functions operate. No new single failure or
accident scenarios are created by the proposed change and the
proposed change does not result in any event previously deemed
incredible being made credible.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
No safety analyses were changed or modified as a result of the
proposed TS changes to reflect STS wording for plants with installed
bypass test capability or for aligning TS Condition entry
requirements. All margins associated with the current safety
analyses acceptance criteria are unaffected. The current safety
analyses remain bounding. The safety systems credited in the safety
analyses will continue to be available to perform their mitigation
functions. The proposed change does not affect the availability or
operability of safety-related systems and components.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The Nuclear Regulatory Commission (NRC) staff has reviewed the
licensee's analysis and, based on this review, it appears that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the amendment requests involve no
significant hazards consideration.
Attorney for licensee: James M. Petro, Jr., Esquire, One Cook
Place, Bridgman, MI 49106
NRC Acting Branch Chief: M. Murphy.
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 amendment request: August 14, 2006.
Description of amendment request: The proposed amendments would
make miscellaneous improvements to the Technical Specifications (TS)
for Prairie Island Nuclear Generating Plant (PINGP) Units 1 and 2. The
proposed amendments would revise TS 1.3, ``Completion Times''; TS
3.1.4, ``Rod Group Alignment Limits''; TS 3.3.7, ``Spent Fuel Pool
Special Ventilation System (SFPSVS) Actuation Instrumentation''; TS
3.7.10, ``Control Room Special Ventilation System (CRSVS)''; and TS
Chapter 4.0, ``Design Features''.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
This license amendment request proposes changes to the Prairie
Island Nuclear Generating Plant Technical Specifications as follows:
Technical Specification 1.3, ``Completion Times'', revise a text
header and add a new text header; Technical Specification 3.1.4,
``Rod Group Alignment Limits'', remove a Surveillance Note which
cross-references another Technical Specification and may cause
confusion; Technical Specification 3.3.7, ``Spent Fuel Pool Special
Ventilation System (SFPSVS) Actuation Instrumentation'', revises the
Modes of Applicability consistent with plant design and the
Technical Specifications for the Spent Fuel Pool Special Ventilation
System, the supported system; Technical Specification 3.7.10,
``Control Room Special Ventilation System (CRSVS)'', revises the
applicability of Condition C and clarifies the requirements of the
Surveillance to verify train filtration flow; and Technical
Specification Chapter 4.0, ``Design Features'', revises Reference 1
to the most recent version of the document.
Revising and adding text headers in Technical Specification 1.3
are administrative changes because the revised document does not
change any basis for the current Technical Specifications. Since
these are administrative changes, they do not involve a significant
increase in the probability or consequences of a previously
evaluated accident. Technical Specification 3.1.4 assures that the
control rod positions are within the limits assumed in the safety
analysis and that the assumed shutdown margin is available when
needed. This license amendment request proposes to remove a Note
from a surveillance requirement that cross-references to Technical
Specification 3.1.7. Removal of this Note does not change plant
operations, testing or maintenance; therefore the proposed change
does not involve a significant increase in the probability of an
accident. Since plant operations, testing and maintenance are not
changed, the proposed changes do not involve a significant increase
in the consequences of an accident previously evaluated.
The Spent Fuel Pool Special Ventilation System filters
radioactive materials in the fuel pool enclosure atmosphere released
following a fuel handling accident. This license amendment request
proposes to revise the Modes and Other Specified Conditions of
Applicability for the actuation instrumentation.
Technical Specification to be consistent with the Modes and
Other Specified Conditions of Applicability in the Technical
Specification for the supported system. The Spent Fuel Pool Special
Ventilation System and its actuation instrumentation are not
accident initiators; therefore, the proposed changes do not affect
the probability of an accident. With the proposed change, the
Technical Specifications will continue to require the system
actuation instrumentation to be operable when irradiated fuel is
moved in the fuel pool enclosure which is also the required
Applicability in the supported system Technical Specification. Since
the instrumentation will be required to actuate the supported system
when it is required to operate, the accident consequences will
continue to be mitigated with this proposed Technical Specification
change. Thus, the proposed Technical Specification change does not
involve a significant increase in the consequences of an accident
previously evaluated.
The Control Room Special Ventilation System provides an enclosed
control room environment from which the plant can be operated
following an uncontrolled release of radioactivity. This system is
not an accident initiator, thus the proposed changes do not increase
the probability of an accident. This license amendment proposes
changes which will: (1) Reduce the time to shut down the plant when
Technical Specification required actions or completion time is not
met; and (2) clarifies surveillance requirements to assure that the
system performs as designed. These changes do not impact the
performance of the system; thus this change does not involve a
significant increase in the consequences of an accident previously
evaluated.
Updating the reference in Technical Specification Chapter 4.0 is
an administrative change because the revised document does not
change any basis for the current Technical Specifications. Since
this is an administrative change, it does not involve a significant
increase in the probability or consequences of a previously
evaluated accident.
The changes proposed in this license amendment do not involve a
significant increase 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.
This license amendment request proposes changes to the Prairie
Island Nuclear Generating Plant Technical Specifications as follows:
Technical Specification 1.3, ``Completion Times'', revise a text
header and add a new text header; Technical Specification 3.1.4,
``Rod Group Alignment Limits'', remove a Surveillance Note which
cross-references another Technical Specification and may cause
confusion; Technical Specification 3.3.7, ``Spent Fuel Pool Special
Ventilation System (SFPSVS) Actuation Instrumentation'', revises the
Modes of Applicability consistent with plant design and the
Technical Specifications for the Spent Fuel Pool Special Ventilation
System, the supported system; Technical Specification 3.7.10,
``Control Room Special Ventilation System (CRSVS)'', revises the
applicability of Condition C and clarifies the requirements of the
Surveillance to verify train filtration flow; and Technical
Specification Chapter 4.0, ``Design Features'',
[[Page 67398]]
revises Reference 1 to the most recent version of the document.
Revising and adding text headers in Technical Specification 1.3
are administrative changes because the revised document does not
change any basis for the current Technical Specifications. Since
these are administrative changes, they do not create the possibility
of a new or different kind of accident.
Removal of a surveillance note from Technical Specification
3.1.4 that cross-references another Technical Specification does not
change any plant operations, maintenance activities or testing
requirements. The Limiting Conditions for Operation will continue to
be met and the proper control rod positions will continue to be
maintained. There are no new failure modes or mechanisms created
through the removal of the Surveillance Requirements Note, nor are
new accident precursors generated by this change. This proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed revision of Modes of Applicability for the Spent
Fuel Pool Special Ventilation System actuation instrumentation makes
operation of the actuation instrumentation consistent with the
Technical Specification requirements for the supported system and
does not change the operation of the supported system for accident
mitigation. The Limiting Conditions for Operation will continue to
be met, no new failure modes or mechanisms are created and no new
accident precursors are generated by this change. This proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
The changes proposed for the Control Room Special Ventilation
System Technical Specifications do not change any the system
operations, maintenance activities or testing requirements. The
Limiting Conditions for Operation will continue to be met, no new
failure modes or mechanisms are created and no new accident
precursors are generated by this change. This proposed change does
not create the possibility of a new or different kind of accident
from any previously evaluated.
Updating the reference in Technical Specification Chapter 4.0 is
an administrative change because the revised document does not
change any basis for the current Technical Specifications. Since
this is an administrative change, it does not create the possibility
of a new or different kind of accident.
The Technical Specification changes proposed in this license
amendment 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.
This license amendment request proposes changes to the Prairie
Island Nuclear Generating Plant Technical Specifications as follows:
Technical Specification 1.3, ``Completion Times'', revise a text
header and add a new text header; Technical Specification 3.1.4,
``Rod Group Alignment Limits'', remove a Surveillance Note which
cross-references another Technical Specification and may cause
confusion; Technical Specification 3.3.7, ``Spent Fuel Pool Special
Ventilation System (SFPSVS) Actuation Instrumentation'', revises the
Modes of Applicability consistent with plant design and the
Technical Specifications for the Spent Fuel Pool Special Ventilation
System, the supported system; Technical Specification 3.7.10,
``Control Room Special Ventilation System (CRSVS)'', revises the
applicability of Condition C and clarifies the requirements of the
Surveillance to verify train filtration flow; and Technical
Specification Chapter 4.0, ``Design Features'', revises Reference 1
to the most recent version of the document.
Revising and adding text headers in Technical Specification 1.3
are administrative changes because the revised document does not
change any basis for the current Technical Specifications. Since
these are administrative changes, they do not involve a significant
reduction in a margin of safety.
Plant operations are required to meet all Technical
Specifications for which the Applicability is met; therefore,
removal of the cross-reference Note from a Technical Specification
3.1.4 surveillance requirement does not change how the plant is
operated and therefore, this change does not involve a significant
reduction in a margin of safety.
Technical Specification 3.3.7 provides requirements for
actuation instrument which supports the operation of the Spent Fuel
Pool Special Ventilation System as required by Technical
Specification 3.7.13. The current Applicability for Technical
Specification 3.3.7 requires the actuation instrumentation to be
operable in Modes which are not required by Technical Specification
3.7.13. This license amendment proposes to make Technical
Specification 3.3.7 Applicability the same as Technical
Specification 3.7.13. This change does not reduce the conditions or
Modes when the Spent Fuel Pool Special Ventilation System will
operate and perform its accident mitigation function; thus this
change does not involve a significant reduction in a margin of
safety.
This license amendment proposes changes to the Control Room
Special Ventilation System Technical Specifications which will: (1)
Reduce the time to shut down the plant when Technical Specification
required actions or completion time is not met; and (2) clarifies
surveillance requirements to assure that the system performs as
designed. The proposed time to shut down the plant is consistent
with other Technical Specifications for shutting down the plant and
allows adequate time for an orderly shut down of the plant; thus
this change does not involve a significant reduction in a margin of
safety. The surveillance requirement clarifications do not reduce
any testing requirements and will continue to demonstrate that the
system can perform its required safety function and satisfy the
Limiting Conditions for Operation. Thus this change does not involve
a significant reduction in a margin of safety.
Updating the reference in Technical Specification Chapter 4.0 is
an administrative change because the revised document does not
change any basis for the current Technical Specifications. Since
this is an administrative change, it does not involve a significant
reduction in a margin of safety.
The Technical Specification changes proposed in this license
amendment 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 requests involve no significant hazards consideration.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Branch Chief: M. Murphy (A).
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant (BFN), Units 2 and 3, Limestone County, Alabama
Date of amendment request: October 26, 2006.
Description of amendment request: The proposed request would revise
the Units 2 and 3 emergency diesel generator (EDG) Technical
Specification (TS) Completion Time (CT) from 14 days to 7 days for
restoration of an inoperable EDG. The current 14-day CT was based on
the assumption that Unit 1 was shut down. The near-term restart of Unit
1 will invalidate this assumption, therefore, the affected CTs are
being returned to their original duration of 7 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed Technical Specification change involve a
significant increase in the probability or consequences of an
accident previously evaluated?
Response: No.
The EDGs are designed as backup alternating current (AC) power
sources in the event of a loss of offsite power. The proposed
restoration of the EDG CT to its original TS duration does not
change the conditions, operating configurations, or minimum amount
of operating equipment assumed in the safety analysis for accident
mitigation. No changes are proposed in the manner in which the EDGs
provide plant protection or which create new modes of plant
operation. Therefore, the proposed amendment does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed Technical Specification change create the
possibility of
[[Page 67399]]
a new or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed amendment does not introduce new equipment which
could create a new or different kind of accident. Existing equipment
will not be operated in any new modes or for purposes different than
it is now utilized. No new external threats, release pathways, or
equipment failure modes are created. Therefore, the implementation
of the proposed amendment will not create a possibility for an
accident of a new or different type than those previously evaluated.
3. Does the proposed Technical Specification change involve a
significant reduction in a margin of safety?
Response: No.
BFN's emergency AC [alternating current] system is designed with
sufficient redundancy such that an EDG may be removed from service
for maintenance or testing. The remaining EDGs are capable of
carrying sufficient electrical loads to satisfy the UFSAR [Updated
Final Safety Analysis Report] requirements for accident mitigation
or unit safe shutdown. The proposed change does not impact the
redundancy or availability requirements of offsite power supplies or
change the ability of the plant to cope with station blackout
events.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: L. Raghavan.
U.S. Department of Transportation (USDOT), United States Maritime
Administration (MARAD), License No. NS-1, Docket No. 50-238, Nuclear
Ship Savannah (NSS)
Date of amendment request: August 7, 2006.
Description of amendment request: The proposed license amendment
would modify the Technical Specification (TS) requirements to prepare
for decommissioning the NSS. Five TS changes are proposed. Three of the
proposed changes are related to allowing the NSS to be berthed at
locations other than the James River Reserve Fleet (JRRF), Newport
News, Virginia. The fourth proposed change eliminates the need to
utilize administrative controls to remove the Containment Vessel (CV)
Entry Shield Plugs to perform activities such as surveys, system
walkdowns and inspections required for developing a detailed
decommissioning plan, schedule and cost estimate.
The fifth proposed change clarifies the TS and eliminates
redundancies, subtle differences and inefficiencies in the current TS
regarding preventing unauthorized access into the Reactor Compartment
and Radiation Control Areas. In addition, MARAD is enhancing the
numbering of the TSs to remove ambiguities that exist in the current
numbering (e.g., TS 2.2 is found on pages 3 and 11 of the current TSs).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Proposed changes (1) Ship's Location, (2) Review and Audit
Committee Membership, (3) Qualification to perform Surveys and
Surveillances, (4) CV Entry Shield Plugs and (5) RC and RCA
Entrances are administrative in nature and do not involve the
modification of any plant equipment or affect basic plant operation.
The NSS's reactor is not operational and the level of
radioactivity in the NSS has significantly decreased from the levels
that existed when the 1976 Possession-only License was issued. No
aspect of any of proposed changes is an initiator of any accident
previously evaluated. Consequently, the probability of an accident
previously evaluated is not significantly increased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(2) Does the change create the possibility of a new or different
kind of accident from any accident evaluated?
Response: No.
Proposed changes (1) Ship's Location, (2) Review and Audit
Committee Membership, (3) Qualification to perform Surveys and
Surveillances, (4) CV Entry Shield Plugs and (5) RC and RCA
Entrances are administrative and do not involve any physical
alteration of plant equipment that was not previously allowed by
Technical Specifications. These proposed changes do not change the
method by which any safety-related system performs its function. As
such, no new or different types of equipment will be installed, and
the basic operation of installed equipment is unchanged. The methods
governing plant operation and testing remain consistent with current
safety analysis assumptions.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
(3) Does the change involve a significant reduction in a margin
of safety?
Response: No.
Proposed changes (1) Ship's Location, (2) Review and Audit
Committee Membership, (3) Qualification to perform Surveys and
Surveillances, (4) CV Entry Shield Plugs and (5) RC and RCA
Entrances are administrative in nature. No margins of safety exist
that are relevant to the ship's defueled and partially dismantled
reactor. As such, there are no changes being made to safety analysis
assumptions, safety limits or safety system settings that would
adversely affect plant safety as a result of the proposed changes.
The proposed changes involve movement of the ship, changes in the
performance of responsibilities and significantly improved
radiological conditions since 1976.
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 upon
the staff's review of the licensee's analysis, as well as the staff's
own evaluation, the staff concludes 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.
Senior Technical Advisor, N.S. Savannah: Erhard W. Koehler, MARAD,
Office of Ship Operations.
NRC Branch Chief: Claudia Craig.
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.
[[Page 67400]]
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].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: September 29, 2005, as
supplemented by letter dated July 5, 2006.
Brief description of amendments: These amendments modified the
Security Plan, Training and Qualification Plan, Safeguards Contingency
Plan, and Independent Spent Fuel Security Program.
Date of issuance: October 31, 2006.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment Nos.: Unit 1-162, Unit 2-162, Unit 3-162.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Operating Licenses for all three units.
Date of initial notice in Federal Register: August 1, 2006 (71 FR
43530).
The July 5, 2006, letter contained the no significant hazards
consideration determination for the September 29, 2005, letter that was
published in the August 1, 2006, notice. The July 5, 2006, supplemental
letter provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the staff's original no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 31, 2006.
No significant hazards consideration comments received: No.
Letter contained the no significant hazards consideration
determination for the September 29, 2005, letter that was published in
the August 1, 2006, notice. The July 5, 2006, supplemental letter
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the staff's original no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 31, 2006.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: January 4, 2006.
Brief description of amendment: The proposed amendment changed the
Millstone Power Station, Unit No. 2 Technical Specification (TS) 3/4
3.3.8, ``Instrumentation, Accident Monitoring,'' to modify the
description of the pressurizer power operated relief valves and
pressurizer safety valves position indicators.
Date of issuance: November 7, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 294.
Facility Operating License No. DPR-65: The amendment revised the
TSs.
Date of initial notice in Federal Register: February 28, 2006 (71
FR 10073).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 7, 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: November 15, 2005.
Brief description of amendment: The amendment modified the
technical specifications to clarify the wording of the emergency closed
cooling water (ECCW) Surveillance Requirement 3.7.10.2 that verified
actuation of the entire ECCW system rather than just verifying
``valve'' actuation.
Date of issuance: October 27, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 139.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specification Surveillance Requirements and License.
Date of initial notice in Federal Register: January 31, 2006 (71 FR
5081).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 27, 2006.
No significant hazards consideration comments received: No.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant, Units 3 and 4, Miami-Dade County, Florida
Date of application for amendments: April 27, 2006, as supplemented
October 3, 2006.
Brief description of amendments: The amendments revise, on a one-
time basis, Technical Specification 3/4.4.5, Steam Generator (SG)
Surveillance Requirements, to exclude the region of the SG tubes below
17 inches from the top of the hot leg tube sheet from the inspection
requirements. The amendments also permanently revise the limit for
primary-to-secondary leakage in TS 3/4.4.6, Reactor Coolant System
Leakage.
Date of issuance: November 1, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos: 231 and 226.
Renewed Facility Operating License Nos. DPR-31 and DPR-41:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 1, 2006 (71 FR
43532).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 1, 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, as supplemented
by letter dated August 3, 2006.
Brief description of amendments: The amendment revised Section
3.3.1, ``Reactor Trip System (RTS) Instrumentation,'' of the DCCNP-1
and DCCNP-2 Technical Specifications, changing the reactor trip on
turbine trip interlock from the P-7 setpoint (10 percent rated thermal
power) to the P-8 setpoint (31 percent rated thermal power).
Date of issuance: October 30, 2006.
[[Page 67401]]
Effective date: As of the date of issuance and shall be implemented
prior to entry into Mode 1 from the Cycle 21 refueling outage for
DCCNP-1, and prior to entry into Mode 1 from the Cycle 17 refueling
outage for DCCNP-2.
Amendment Nos.: 297 and 298.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: April 25, 2006 (71 FR
23956).
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 October 30, 2006.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: March 15, 2006.
Brief description of amendment: The amendment revised the Cooper
Nuclear Station Technical Specification 5.5.12, ``Primary Containment
Leakage Rate Testing Program,'' by adding two sub-paragraphs to note
exemptions from Section III.A and Section llI.B of 10 CFR Part 50,
Appendix J, Option B. These two sub-paragraphs allow the leakage
contribution from the four main steam line penetrations, referred to as
the Main Steam Isolation Valve leakage, to be excluded.
Date of issuance: October 31, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 226.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 25, 2006 (71 FR
23958).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 31, 2006.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1 (FCS), Washington County, Nebraska
Date of amendment request: October 31, 2005, as supplemented on
July 25, 2006.
Brief description of amendment: The amendment revised the FCS
Updated Safety Analysis Report Sections related to the radiological
consequences of events affected by the planned 2006 replacement of the
steam generators and pressurizer.
Date of issuance: October 27, 2006.
Effective date: As of its date of issuance and shall be implemented
within 90 days of its issuance.
Amendment No.: 243.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Updated Safety Analysis Report.
Date of initial notice in Federal Register: December 20, 2005 (70
FR 75493).
The July 25, 2006, supplemental letter provided information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated October 27, 2006.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: December 19, 2005, as supplemented on
May 30, 2006.
Brief description of amendment: The amendment modified Fort Calhoun
Station, Unit No. 1's Technical Specification 2.4, ``Containment
Cooling,'' (and the associated Bases) to reduce the required number of
operable containment spray (CS) pumps from three to two in order to
enhance net positive suction head margins. The proposed change was
implemented by disabling the CS actuation signal automatic start
feature of one of the two CS pumps that share the same diesel generator
and a common suction line.
Date of issuance: October 27, 2006.
Effective date: The license amendment is effective as of its date
of issuance.
Amendment No.: 244.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: February 28, 2006 (71
FR 10075).
The May 30, 2006, supplemental letter provided information that
clarified the application, did not expand the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated October 27, 2006.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: September 30, 2005, as supplemented by
letters dated May 23 and August 16, 2006.
Brief description of amendment: Omaha Public Power District
proposed to change the licensing basis by replacing EMF-2087(P)(A),
Revision 0, ``SEM/PWR-98: ECCS [Emergency Core Cooling System]
Evaluation Model for PWR [Pressurized-Water Reactor] LBLOCA [Large
Break Loss-of-Coolant Accident] Applications,'' Siemens Power
Corporation, June 1999, with the AREVA NP, Inc. Topical Report EMF-
2103(P)(A), ``Realistic Large Break LOCA Methodology,'' Framatome ANP,
Inc., in the Fort Calhoun Station, Unit 1 (FCS), Core Operating Limit
Report (COLR). This change is necessary since the EMF-2087(P)(A)
methodology is not approved for analyzing M5 clad fuel, which will be
used in the FCS reactor core starting in Cycle 24. As part of this
approval, the NRC staff reviewed the AREVA NP, Inc. FCS-specific LBLOCA
analysis using EMF-2103(P)(A). EMF-2103(P)(A) will be used for Cycle 24
and beyond.
Date of issuance: November 3, 2006.
Effective date: Effective as of its date of issuance and shall be
implemented within 90 days of issuance.
Amendment No.: 245.
Renewed Facility Operating License No. DPR-40: The amendment
revised the COLR.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
152).
The May 23 and August 16, 2006, supplemental letters provided
additional information that clarified the application, did not expand
the scope of the application as originally noticed, and did not change
the staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a safety evaluation dated November 3, 2006.
No significant hazards consideration comments received: No.
Omaha Public Power District (OPPD), Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: May 30, 2006, as supplemented by two
letters dated on August 30, 2006.
[[Page 67402]]
Brief description of amendment: The amendment revised the Fort
Calhoun Station, Unit No. 1 (FCS) Technical Specification (TS)
requirements related to steam generator tube integrity. The change is
consistent with NRC-approved Revision 4 to Technical Specification Task
Force (TSTF) Standard Technical Specification Change Traveler TSTF-449,
``Steam Generator Tube Integrity.'' The availability of this TS
improvement was announced in the Federal Register on May 6, 2005 (70 FR
24126) as part of the consolidated line item improvement process
(CLIIP).
OPPD also changed the FCS TS by deleting the sleeving repair
alternative to plugging for steam generator tubes. The FCS replacement
steam generators (RSGs) to be installed during the fall of 2006 are
manufactured by Mitsubishi Heavy Industries, Ltd. (MHI). OPPD has
stated that the sleeving repair alternative to plugging will not be
used for the MHI RSGs.
Date of issuance: November 7, 2006.
Effective date: As of its date of issuance and shall be implemented
within 120 days of issuance.
Amendment No.: 246.
Renewed Facility Operating License No. DPR-40: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40750).
The two August 30, 2006, supplemental letters provided information
that clarified the application, did not expand the scope of the
application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
safety evaluation dated November 7, 2006.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of application for amendment: October 7, 2005, as supplemented
by letter dated September 8, 2006.
Brief description of amendment: The proposed amendment revised the
Technical Specifications (TSs) to clarify certain requirements during
fuel movement, core alterations, and operations with the potential for
draining the reactor vessel. The amendment better aligns the TSs with
the NRC-approved Revision 2 to Technical Specification Task Force
(TSTF) Traveler TSTF-51, ``Revise Containment Requirements During
Handling Irradiated Fuel and Core Alterations,'' and NUREG-1433,
``Standard Technical Specifications General Electric Plants, BWR
[boiling water reactor]/4.''
Date of issuance: October 31, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 170.
Facility Operating License No. NPF-57: This amendment revised the
TSs.
Date of initial notice in Federal Register: May 9, 2006 (71 FR
27002).
The licensee's September 8, 2006, supplement provided clarifying
information that did not change the scope of the proposed amendment as
described in the original notice of proposed action published in the
Federal Register, and did not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 31, 2006.
No significant hazards consideration comments received: No.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of application for amendments: April 25, 2006.
Brief description of amendments: The amendments revised the
Technical Specifications to adopt the provisions in Technical
Specification Task Force (TSTF) Traveler TSTF-359, ``Increased
Flexibility in Mode Restraints,'' Revision 9. The availability of TSTF-
359 for adoption by licensees was announced in the Federal Register on
April 4, 2003 (68 FR 16579).
Date of issuance: October 27, 2006.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos.: 276, 258.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications and License.
Date of initial notice in Federal Register: July 5, 2006 (71 FR
38185).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 27, 2006.
No significant hazards consideration comments received: No.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: October 28, 2005, as
supplemented on April 2, June 15, and August 31, 2006.
Brief description of amendment: The amendment revises the Virgil C.
Summer Nuclear Station Technical Specifications and provides associated
Bases to permit the implementation of an alternate alternating current
power supply.
Date of issuance: November 2, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No. 178.
Renewed Facility Operating License No. NPF-12: Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: March 14, 2006 (71 FR
13176).
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.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 2, 2006.
No significant hazards consideration comments received: No.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: October 27, 2005.
Brief description of amendment: The amendment revised Technical
Specifications (TSs) 1.1, ``Definitions,'' and 3.4.16, ``RCS [reactor
coolant system] Specific Activity,'' to replace the current Limiting
Condition for Operation (LCO) 3.4.16 limits on RCS specific activity
with limits on RCS Dose Equivalent I-131 (DEI) and Dose Equivalent Xe-
133 (DEX). In TS 1.1, the definition of (1) [Eacute]--Average
Disintegration Energy is replaced by the definition of DEX and (2) DEI
is revised to allow the use of alternate thyroid dose conversion
factors. The modes of applicability, conditions and required actions,
and surveillance requirements for TS 3.4.16 are revised.
Date of issuance: October 31, 2006.
Effective date: As of its date of issuance and shall be implemented
within 90 days of the date of issuance.
Amendment No.: 170.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
156).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 31, 2006.
No significant hazards consideration comments received: No.
[[Page 67403]]
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: August 25, 2006, as supplemented by
letter dated October 25, 2006.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.7.2, ``Main Steam Isolation Valves (MSIVs),'' and
TS 3.7.3, ``Main Feedwater Isolation Valves (MFIVs),'' to add the
associated actuator trains to (1) the limiting condition for operation
(LCO), (2) the conditions, required actions, and completion times for
the LCO, and (3) the surveillance requirements. The Table of Contents
for the TSs is changed to account for the resulting renumbering of TS
pages.
Date of issuance: November 7, 2006.
Effective date: As of its date of issuance and shall be implemented
within 30 days of the date of issuance.
Amendment No.: 171.
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 1, 2006 (71
FR 52173).
The supplemental letter dated October 25, 2006, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the NRC
staff's original proposed no significant hazards consideration
determination published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 7, 2006.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 9th day of November, 2006.
For The Nuclear Regulatory Commission.
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E6-19434 Filed 11-20-06; 8:45 am]
BILLING CODE 7590-01-P