[Federal Register Volume 72, Number 88 (Tuesday, May 8, 2007)]
[Notices]
[Pages 26173-26184]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-8679]
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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 April 13, 2007 to April 26, 2007. The last
biweekly notice was published on April 24, 2007 (72 FR 20375).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The
[[Page 26174]]
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: November 27, 2006.
Description of amendment request: The amendment would revise the
Oyster Creek Technical Specification (TS) 6.9.1.d, ``Annual Radioactive
Effluent Release Report,'' by changing the requirement to submit the
report within 60 days of January 1. Specifically, the revised
requirement would be to submit the report prior to May 1 of each year.
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 involves a revision to the required
submittal date for the Radioactive Effluent Release Report, and is
administrative in nature. The change will not alter the physical
design or operation of any plant structure, system, or component.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The proposed change is administrative in nature. The proposed
change has no impact on the design, function or operation of any
plant structure, system or component and does not affect any
accident analyses. Accordingly, the change does not introduce any
new accident initiators, nor does it reduce or adversely affect the
capabilities of any plant structure, system, or component to perform
their safety function.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change is administrative in nature, does not negate
any existing requirement, and does not adversely affect existing
plant safety margins or the reliability of the equipment assumed to
operate in the safety analysis. As such, there is no change being
made to safety analysis assumptions, safety limits or safety system
settings that would adversely affect plant 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: Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LCC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
[[Page 26175]]
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: March 28, 2007.
Description of amendment request: The proposed change would revise
the required wattage specified in the River Bend Station, Unit 1 (RBS),
Technical Specification 5.5.7.e, Ventilation Filter Testing Program,
for the Control Room Fresh Air System (CRFAS) heater for testing. The
proposed required wattage for testing the CRFAS heater would be revised
from 23 2.3 kilowatt (kW), to ``>==15 kW.''
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.
This change specifies the required power (in kW) for the Control
Room ventilation electric heaters to decrease relative humidity of
the air to less than 70% relative humidity as required for proper
operation of the charcoal absorber components based on calculated
requirements. The heater will continue to perform its intended
design function as designed. The heater is not an accident
precursor.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The heater will continue to perform its function as designed.
The heater provides humidity control for the Control Room filter
unit during a design basis accident. Changing the test acceptance
criteria to a calculated value has no influence on, nor does it
contribute in any way to, the possibility of a new or different kind
of accident or malfunction from those previously analyzed. No change
has been made to the design, function or method of performing
testing. No safety-related equipment or safety functions are altered
as a result of this change.
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 margin of safety is changed as a result of this change. The
heater will continue to perform its design function. Testing
methodology has not changed. The function of the heater is
unchanged. The acceptance criterion has been changed to a calculated
value rather than the name plate rating to make testing more
realistic. The heater will continue to operate to perform its
intended design function.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: March 30, 2007.
Description of amendment request: The proposed amendment would
revise Arkansas Nuclear One, Unit 2 (ANO-2) Technical Specification
(TS) to support a partial re-rack of the storage racks in the ANO-2
spent fuel pool (SFP). The proposed amendment would revise TS 3.9.12,
``Fuel Storage,'' and its associated tables, figures, and surveillance
requirements, TS 5.3, ``Fuel Storage,'' and add TS 6.5.17, ``Metamic
Coupon Sampling Program.'' The ANO-2 TS 3.9.12 would be changed to: (1)
Support higher fuel assembly U-235 enrichment; (2) apply the
appropriate loading restrictions; and (3) delete the dry cask loading
restrictions. ANO-2 TS 5.3.1b would be changed to reflect a different
SFP boron concentration that is needed to assure K-effective
(Keff) remains less than or equal to 0.95. ANO-2 TS 5.3.2a
would be modified to reflect a higher fuel assembly U-235 enrichment. A
new coupon sampling program would be added as TS 6.5.17. In addition,
TS Surveillance Requirement 4.9.12.d would be added to direct
performance of the coupon sampling program.
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.
Fuel Handling Accidents
The current licensing bases for the dose consequences associated
with a fuel handling accident (FHA), which was performed considering
a maximum U-235 enrichment of 5.0 wt% and a maximum burnup of 65
megawatt-days/kilograms of uranium, does not exceed 25% of 10 CFR
[Title 10 of the Code of Federal Regulations] 100 limits. The
proposed change is bounded by the current analysis and therefore,
there is no increase in the dose consequences associated with a[n]
FHA.
During rack removal and installation, safe load paths will be
determined and written procedures followed to ensure that the racks
are not carried over any fuel assemblies. With the proposed
limitations on rack and cask movement, there should be no impact to
spent fuel and no radiological consequences due to fuel rack
installation. The racks will be moved with a single failure proof
crane. Therefore, a postulated drop of a rack is not a credible
accident.
The probability of having a[n] FHA has not increased.
Criticality Accidents Associated With a Dropped Fuel Assembly
The three fuel assembly drop accidents described below can be
postulated to increase reactivity. However, for these accident
conditions, the double contingency principle of ANS [American
National Standard] N16.1-1975 is applied. This states that it is
unnecessary to assume two unlikely, independent, concurrent events
to ensure protection against a criticality accident. Thus, for
accident conditions, the presence of soluble boron in the storage
pool water can be assumed as a realistic initial condition since its
absence would be a second unlikely event.
Three types of drop accidents have been considered: a vertical
drop accident, a horizontal drop accident, and an inadvertent drop
of an assembly between the outside periphery of the rack and the
pool wall. The structural damage to the pool liner, the racks, and
fuel assembly resulting from a dropped fuel assembly striking the
rack, the pool floor, or another assembly located in the racks is
primarily dependent on the mass of the falling object, drop height,
and structural configuration of the rack. The two parameters related
to the fuel assembly (mass and drop height) are not changed by the
proposed rack modification. The new rack design was evaluated for
all postulated structural drops and the structural damage to these
items remains within acceptable limits. In all cases the proposed TS
limit for boron concentration ensures that a five percent
subcriticality margin is met for the postulated accidents.
Criticality Accidents Associated With a Misplaced Fuel Assembly
The fuel assembly misplacement accident was considered for all
storage configurations. An assembly with high reactivity is assumed
to be placed in a storage location which requires restricted storage
based on initial U-235 loading, cooling time, and burnup. The
presence of boron in the pool water assumed in the analysis has been
shown to offset the worst case reactivity effect of a misplaced
[[Page 26176]]
fuel assembly for any configuration. This boron requirement is less
than the boron concentration required by the ANO-2 TS. Thus, a five
percent subcriticality margin is met for postulated accidents, since
any reactivity increase will be much less than the negative worth of
the dissolved boron.
Optimum Moderation Accident
For fuel storage applications in the SFP, water is usually
present. An ``optimum moderation'' accident is not a concern in SFP
storage racks because the rack design prevents the preferential
reduction of water density between the cells of a rack (e.g.,
boiling between cells). In addition, the criticality analysis has
demonstrated that the effective neutron multiplication factor
(Keff) will remain less than 1.0 when the SFP is fully
flooded with unborated water.
An ``optimum moderation'' accident in the new fuel vault was
evaluated and the conclusions of that evaluation confirmed that the
reactivity effect is less than the regulatory limit of 0.98 for
Keff.
Loss of SFP Cooling
The proposed modification to the ANO-2 SFP racks does not result
in a change to the SFP cooling system and therefore the probability
of a loss of SFP cooling is not increased.
The consequences of a loss of spent fuel pool cooling were
evaluated and found to not involve a significant increase as a
result of the proposed changes. A thermal-hydraulic evaluation for
the loss of SFP cooling was performed. The analysis determined that
the minimum time to boil is about two hours following a complete
core off load and a complete loss of forced cooling. This provides
sufficient time for the operators to restore cooling or establish an
alternate means of cooling before the water shielding above the top
of the racks falls below 10 feet. Therefore, the proposed change
represents no increase in the consequences of loss of pool cooling.
Seismic Event
The proposed rack modification does not result in an increase in
the probability or consequences of a design basis seismic event. The
new racks were analyzed and all structural acceptance criteria are
shown to be met during seismic events. The structural capability of
the SFP and liner will not be exceeded as a result of the new rack
design.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The presence of soluble boron in the pool water assumed in the
criticality analysis is less than the boron concentration required
by the ANO-2 TSs. Thus, a five percent subcriticality margin is met
for postulated accidents, since any reactivity increase will be much
less than the negative worth of the dissolved boron.
No new or different types of fuel assembly drop scenarios are
created by the proposed change. During the installation of the new
racks, the possibility of dropping a rack is not a credible accident
since a single failure proof crane and safe load paths will be used
for rack movements. No new or different fuel assembly misplacement
accidents will be created. Administrative controls currently exist
to assist in assuring fuel misplacement does not occur.
No changes are proposed to the spent fuel pool cooling system or
makeup systems and therefore no new accidents are considered related
to the loss of cooling or makeup capability.
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.
With the presence of a nominal boron concentration, the SFP
storage racks will be designed to assure a subcritical array with a
five percent subcritical margin (95% probability at the 95 %
confidence level). This has been verified by criticality analyses.
Credit for soluble boron in the SFP water is permitted under
accident conditions. The proposed modification that will allow
installation of the new racks does not result in the potential of
any new misplacement scenarios. Criticality analyses have been
performed to determine the required boron concentration that would
ensure the maximum Keff does not exceed 0.95. The ANO-2
TS for the minimum SFP boron concentration is greater than that
required to ensure Keff remains below 0.95. Therefore,
the margin of safety defined by taking credit for soluble boron will
be maintained.
The structural analysis of the new spent fuel racks along with
the evaluation of the SFP structure indicated that the integrity of
these structures will be maintained. The structural requirements
were shown to be satisfied, thus the safety margins were maintained.
In addition the proposed change includes a coupon sampling
program that will monitor the physical properties of the
MetamicTM absorber material. The monitoring program
provides a method of verifying that the assumptions used in the SFP
criticality analyses remain valid.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of amendment request: March 1, 2007.
Description of amendment request: The proposed change would revise
Grand Gulf Nuclear Station, Unit 1, Technical Specification (TS) Tables
3.3.5.1-1 and 3.3.5.2-1 to modify the allowable values of the low
Condensate Storage Tank (CST) level setpoints for the High Pressure
Core Spray (HPCS) and Reactor Core Isolation Cooling (RCIC) suction
swap from the CST to the Suppression Pool. The change is necessary to
correct an error in the original plant design. The error, under certain
conditions, could prevent a swap of the HPCS and RCIC suction flow
paths to the Suppression Pool. Currently, the erroneous setpoints have
been corrected to a higher level, and are administratively controlled
in accordance with the Administrative Letter 98-10, ``Dispositioning of
Technical Specifications That Are Insufficient To Assure Plant
Safety.''
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.
This change will adjust the setpoint for an automatic swap of
the suction for the HPCS and RClC systems from the Condensate
Storage Tank (CST) to the Suppression Pool. The Suppression Pool is
the source of water credited in the accident analyses. This transfer
is not the initiator of any analyzed accident. The setpoint
adjustment will allow a transfer of the suction to an assured
safety-related water source earlier in the event and will have no
effect on the probability or consequences of an accident previously
evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Transfer of the suction source for HPCS and RClC will occur
sooner as a result of this change. No new operational conditions
beyond those currently allowed are introduced. This change is
consistent with the safety analyses assumptions and current
[[Page 26177]]
plant operating practices. This simply corrects the setpoint
consistent with the accident analyses and therefore cannot create
the possibility of a new or different kind of accident from any
previously evaluated accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not reduce safety, but rather allows
the transfer from the CST to the Suppression Pool sooner. The
Suppression Pool is the source of water credited in the accident
analyses. This change is consistent with the safety analyses
assumptions and current plant operating practices. No new
operational conditions beyond those currently allowed are created by
these changes.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: October 18, 2006.
Description of amendment request: A change is proposed to the
technical specifications (TSs) of LaSalle County Station, Units 1 and 2
(LaSalle), consistent with TS Task Force Traveler No. 432 (TSTF-423),
``Technical Specification End States, NEDC-32988-A,'' to the standard
TSs for boiling-water reactor plants, to allow for some systems entry
into hot shutdown rather than cold shutdown, to repair equipment if
risk is assessed and managed consistent with the program in place for
complying with the requirements of Title 10 of the Code of Federal
Regulations (10 CFR) Section 50.65(a)(4). The proposed amendment would
modify the TS to risk-informed requirements regarding selected required
action end states provided in TSTF-423, Revision 0.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1: The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows a change to certain required end
states when the TS Completion Times for remaining in power operation
will be exceeded. Most of the requested technical specification (TS)
changes are to permit an end state of hot shutdown (Mode 3) rather
than an end state of cold shutdown (Mode 4) contained in the current
TS. The request was limited to: (1) Those end states where entry
into the shutdown mode is for a short interval, (2) entry is
initiated by inoperability of a single train of equipment or a
restriction on a plant operational parameter, unless otherwise
stated in the applicable technical specification, and (3) the
primary purpose is to correct the initiating condition and return to
power operation as soon as is practical. Risk insights from both the
qualitative and quantitative risk assessments were used in specific
TS assessments. Such assessments are documented in Section 6 of GE
NEDC-32988, Revision 2, ``Technical Justification to Support Risk
Informed Modification to Selected Required Action End States for BWR
Plants.'' They [risk assessments] provide an integrated discussion
of deterministic and probabilistic issues, focusing on specific
technical specifications, which are used to support the proposed TS
end state and associated restrictions. The [NRC] staff finds that
the risk insights support the conclusions of the specific TS
assessments. Therefore, the probability of an accident previously
evaluated is not significantly increased, if at all. The
consequences of an accident after adopting proposed TSTF-423, are no
different than the consequences of an accident prior to adopting
TSTF-423. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2: The Proposed Change Does Not Create The Possibility of a
New or Different Kind of Accident From Any Previously Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
If risk is assessed and managed, allowing a change to certain
required end states when the TS Completion Times for remaining in
power operation are exceeded, i.e., entry into hot shutdown rather
than cold shutdown to repair equipment, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change and the commitment by the licensee to adhere to the guidance
in TSTF-IG-05-02, Implementation Guidance for TSTF-423, Revision 0,
``Technical Specifications End States, NEDC-32988-A,'' will further
minimize possible concerns. Thus, this change does not create the
possibility of a new or different kind of accident from an accident
previously evaluated.
Criterion 3: The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety.
The proposed change allows, for some systems, entry into hot
shutdown rather than cold shutdown to repair equipment, if risk is
assessed and managed. The [Boiling Water Reactor Owners Group]
BWROG's risk assessment approach is comprehensive and follows [NRC]
staff guidance as documented in [Regulatory Guides] RGs 1.174 and
1.177. In addition, the analyses shows that the criteria of the
three-tiered approach for allowing TS changes are met. The risk
impact of the proposed TS changes was assessed following the three-
tiered approach recommended in RG 1.177. A risk assessment was
performed to justify the proposed TS changes. The net change to the
margin of safety is insignificant. Therefore, this change does not
involve a significant reduction in a margin of safety.
LaSalle has reviewed the proposed no significant hazards
consideration determination published on March 23, 2006, (71 FR 14743)
as part of the consolidated line item improvement 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, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
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: February 28, 2007.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) 5.2.2, ``Plant Staff'', and TS 5.3,
``Plant Staff Qualifications'', requirements for shift technical
advisor (STA) qualifications. The proposed changes will specify that
personnel who perform the function of STA shall meet the qualification
requirements of the Commission Policy Statement on Engineering
Expertise on Shift, published in the Federal Register on October 28,
1985 (50 FR 43621). This change will allow qualified personnel to
perform the function of STA without
[[Page 26178]]
also holding a senior reactor operator (SRO) license.
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 to add a new sentence to
Technical Specification 5.2.2 specifying that personnel who perform
the function of shift technical advisor shall meet the qualification
requirements of the Commission Policy Statement on Engineering
Expertise on Shift and remove shift technical advisor qualification
requirements from Technical Specification 5.3.1. This change will
allow qualified personnel to perform the function of shift technical
advisor without also holding a senior reactor operator license.
The proposed changes are administrative changes to Technical
Specifications Chapter 5, the administrative chapter of the
Technical Specifications. Shift technical advisors perform the
function of on-shift technical advisor to the shift supervisor and
do not operate the plant. Therefore, the changes proposed in this
license amendment request do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
This license amendment request proposes to add a new sentence to
Technical Specification 5.2.2 specifying that personnel who perform
the function of shift technical advisor shall meet the qualification
requirements of the Commission Policy Statement on Engineering
Expertise on Shift and remove shift technical advisor qualification
requirements from Technical Specification 5.3.1. This change will
allow qualified personnel to perform the function of shift technical
advisor without also holding a senior reactor operator license.
The Technical Specification changes proposed in this license
amendment are administrative, do not change the manner in which the
plant is operated, and 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 to add a new sentence to
Technical Specification 5.2.2 specifying that personnel who perform
the function of shift technical advisor shall meet the qualification
requirements of the Commission Policy Statement on Engineering
Expertise on Shift and remove shift technical advisor qualification
requirements from Technical Specification 5.3.1. This change will
allow qualified personnel to perform the function of shift technical
advisor without also holding a senior reactor operator license.
The proposed changes are administrative changes to Technical
Specifications Chapter 5, the administrative chapter of the
Technical Specifications. Shift technical advisors perform the
function of on-shift technical advisor to the shift supervisor and
do not operate the plant. Thus, the Technical Specification changes
proposed in this license amendment request 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: L. Raghavan.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: January 18, 2007.
Brief description of amendments: The amendments requested would
revise Technical Specifications (TS) requirement 3.8.1, ``AC Sources--
Operating,'' Extension of Completion Times for Offsite Circuits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed Technical Specification (TS) Completion Time (CT)
extension does not significantly increase the probability of
occurrence of a previously evaluated accident because the startup
transformers (STs) are not initiators of previously evaluated
accidents involving a loss of offsite power (LOOP). The proposed
changes to the TS Required Actions CTs do not affect any of the
assumptions used in the deterministic or the PSA [probabilistic
safety assessment] analysis relative to LOOP initiating event
frequency. Implementation of the proposed changes does not result in
a risk significant impact. The onsite AC [alternating current] power
sources will remain highly reliable and the proposed changes will
not result in a significant increase in the risk of plant operation.
This is demonstrated by showing that the impact on plant safety as
measured by the increase in core damage frequency (CDF) is less than
1E-06 per year and the increase in large early release frequency
(LERF) is less than 1E-07 per year. In addition, for the CT changes,
the incremental conditional core damage probabilities (ICCDP) and
incremental conditional large early release probabilities (ICLERP)
are less than 5E-07 and 5E-08, respectively. These changes meet the
acceptance criteria in Regulatory Guides 1.174 and 1.177. Therefore,
since the onsite AC power sources will continue to perform their
functions with high reliability as originally assumed and the
increase in risk as measured by [Delta]CDF, [Delta]LERF, ICCDP, and
ICLERP risk metrics is within the acceptance criteria of existing
regulatory guidance, there will not be a significant increase in the
consequences of any accidents.
The proposed changes do not adversely affect accident initiators
or precursors nor alter the design assumptions, conditions, or
configuration of the facility or the manner in which the plant is
operated and maintained. The proposed changes do not alter or
prevent the ability of structures, systems, and components (SSCs)
from performing their intended function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
The proposed changes are consistent with safety analysis assumptions
and resultant consequences.
The proposed TS CT extension will continue to provide assurance
that the sources of power to 6.9 kV [kilovolts] AC buses perform
their function when called upon. Extending the TS CT to 30 days does
not affect the design of the STs, the operational characteristics of
the STs, the interfaces between the STs and other plant systems, the
function, or the reliability of the STs. Thus, the STs will be
capable of performing their accident mitigation functions and there
is no impact to the radiological consequences of any accident
analysis.
The Configuration Risk Management Program (CRMP) in TS 5.5.18 is
an administrative program that assesses risk based on plant status.
The risk-informed CT will be implemented consistent with the CRMP
and approved plant procedures. When utilizing the 30-day extension,
requirements of the CRMP per TS 5.5.18 call for the consideration of
other measures to mitigate the consequences of an accident occurring
while a[n] ST is inoperable. Furthermore, administrative controls
will be applied when exercising the 30-day CT extension and are
adequate to maintain defense-in-depth and sufficient safety margins.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
[[Page 26179]]
Response: No.
The proposed changes do not result in a change in the manner in
which the electrical distribution subsystems provide plant
protection. There [are] no design changes associated with the
proposed changes. The changes to the CT do not change any existing
accident scenarios, nor create any new or different accident
scenarios.
The changes do not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. In addition,
the changes do not impose any new or different requirements or
eliminate any existing requirements. The changes do not alter any of
the assumptions made in the safety analysis. The changes to the CT
do not affect the accident analysis directly; the CT is strictly
tied to the PRA [probabilistic risk assessment] and the risk
associated with the occurrence of a low-probability event during the
limited time the component is unavailable.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not alter the manner in which safety
limits, limiting safety system settings or limiting conditions for
operation are determined. Neither the safety analyses nor the safety
analysis acceptance criteria are impacted by these changes. The
proposed changes will not result in plant operation in a
configuration outside the current design basis. The proposed
activities only involve changes to certain TS CTs.
The proposed change does not involve a change to the plant
design or operation and thus does not affect the design of the STs,
the operation characteristics of the STs, the interfaces between the
STs and other plant systems, or the function or reliability of the
STs. Because the STs' performance and reliability will continue to
be ensured by the proposed TS change, the proposed changes do not
result in a reduction in the margin of safety.
Therefore the proposed change does not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: David Terao.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: December 19, 2006.
Brief description of amendments: The amendments requested would
revise Technical Specification (TS) requirement 5.5.16, ``Containment
Leakage Rate Testing Program,'' for consistency with the requirements
of paragraph 50.55a(g)(4) of Title 10 of the Code of Federal
Regulations (10 CFR) for components classified as Code Class CC.
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[es] the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the TS administrative controls
programs for consistency with the requirements of 10 CFR [Part] 50,
paragraph 55a(g)(4) for components classified as Code Class CC.
The proposed change affects the frequency of visual examinations
that will be performed for the concrete surfaces of the containment
for the purpose of the Containment Leakage Rate Testing Program. In
addition, the proposed change allows those examinations to be
performed during power operation as opposed to during a refueling
outage. The frequency of visual examinations of the concrete
surfaces of the containment and the mode of operation during which
those examinations are performed has no relationship to or adverse
impact on the probability of any of the initiating events assumed in
the accident analyses. The proposed change would allow visual
examinations that are performed pursuant to NRC approved [American
Society of Mechanical Engineers] (ASME) Section XI Code requirements
(except where relief has been granted by the NRC) to meet the intent
of visual examinations required by Regulatory Guide 1.163, without
requiring additional visual examinations pursuant to the Regulatory
Guide. The intent of early detection of deterioration will continue
to be met by the more rigorous requirements of the Code required
visual examinations. As such, the safety function of the containment
as a fission product barrier is maintained.
The proposed change does not impact any accident initiators or
analyzed events or assumed mitigation of accident or transient
events. It does not involve the addition or removal of any
equipment, or any design changes to the facility.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do[es] the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change revises the TS administrative controls
programs for consistency with the requirements of 10 CFR [Part] 50,
paragraph 55a(g)(4) for components classified as Code Class CC.
The change affects the frequency of visual examinations that
will be performed for the concrete surfaces containments. In
addition, the proposed change allows those examinations to be
performed during power operation as opposed to during a refueling
outage. The proposed change does not involve a modification to the
physical configuration of the plant (i.e., no new equipment will be
installed) or change in the methods governing normal plant
operation. The proposed change will not impose any new or different
requirements or introduce a new accident initiator, accident
precursor, or malfunction mechanism. Additionally, there is no
change in the types or increases in the amounts of any effluent that
may be released off-site and there is no increase in individual or
cumulative occupational exposure.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Do[es] the proposed change involve a significant reduction in
a margin of safety?
Response: No.
The proposed change revises the Improved Standard Technical
Specification Administrative Controls program requirements for
consistency with the requirements of 10 CFR [Part] 50, paragraph
55a(g)(4) for components classified as Code Class CC.
The change affects the frequency of visual examinations that
will be performed for the concrete surfaces containments. In
addition, the proposed change allows those examinations to be
performed during power operation as opposed to during a refueling
outage. The safety function of the containment as a fission product
barrier will be maintained.
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: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Branch Chief: Thomas Hiltz.
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
[[Page 26180]]
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 51.22(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of application for amendment: March 28, 2005, as supplemented
by letters dated November 2, 2005, January 24, February 2, March 16,
March 23, and March 28, 2007.
Brief description of amendment: The amendment revises the Oyster
Creek Licensing Basis in the area of radiological dose analyses for
design-basis accidents using the alternative source terms depicted in
Regulatory Guide 1.183, ``Alternative Radiological Source Terms for
Evaluating Design Basis Accidents at Nuclear Power Reactors.''
Additionally, the amendment revises the Oyster Creek Technical
Specifications (TSs) consistent with the amended design-basis.
Date of Issuance: April 26, 2007.
Effective date: As of the date of Issuance to be implemented within
60 days.
Amendment No.: 262.
Facility Operating License No. DPR-16: The amendment revised the
TSs.
Date of initial notice in Federal Register: May 10, 2005 (70 FR
24646). The 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 Nuclear Regulatory
Commission (NRC) staff's original proposed to significant hazards
consideration determination.
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated April 26, 2007.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2 (HBRSEP2), Darlington County, South Carolina
Date of application for amendment: June 1, 2006, as supplemented by
letters dated November 20, 2006, and February 22, 2007.
Brief description of amendment: The amendment revises Surveillance
Requirement 3.5.2 in the HBRSEP2 Technical Specifications.
Date of issuance: April 4, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 213.
Renewed Facility Operating License No. DPR-23: Amendment revises
the Technical Specifications.
Date of initial notice in Federal Register: December 19, 2006 (71
FR 75992). The supplemental letters provided additional information
that was within the scope of the original notice and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 4, 2007.
No significant hazards consideration comments received: No.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: November 27, 2006.
Brief description of amendment: The amendment revised Technical
Specification (TS) 5.5.9 to relocate the specific American Society of
Testing and Materials (ASTM) Standard from the Administrative Controls
Section of TS to a licensee-controlled document. Also, the revision to
TS 5.5.9 allows the performance of an alternate water and sediment
content test to establish the acceptability of new fuel oil prior to
addition to the storage tank has been added to the clear and bright
test.
Date of issuance: April 12, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 174.
Facility Operating License No. NPF-43: Amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: January 3, 2007 (72 FR
149).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 12, 2007.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: October 13, 2006.
Brief description of amendment: The amendment revised Facility
Operating License No. NPF-58 by deleting License Condition 2.F, which
specifies reporting of 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).
Date of issuance: April 19, 2007.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 140.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: November 21, 2006 (71
FR 67394).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 19, 2007.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of application for amendment: August 31, 2006, as supplemented
on December 15, 2006, and March 1 and April 4, 2007.
Brief description of amendment: The amendment conforms the license
to reflect the transfer of Renewed Facility Operating License No. DPR-
20 to Entergy Nuclear Palisades, LLC, as
[[Page 26181]]
owner, and Entergy Nuclear Operations, Inc., as operator, as approved
by Order of the Commission dated April 6, 2007, and as revised on April
10, 2007.
Date of issuance: April 11, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 224.
Facility Operating License No. DPR-20: Amendment revised the
Renewed Operating License and Technical Specifications.
Date of initial notice in Federal Register: November 16, 2006 (71
FR 66805).
The December 15, 2006, and March 1 and April 4, 2007, supplemental
letters contained clarifying information 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 April 6, 2007, as revised on April 10,
2007.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power
Plant, Unit 3, Humboldt County, California
Date of application for amendment: December 20, 2006.
Brief description of amendment: The amendment revises License
Condition 2.B.3(c) to allow the receipt, possession, and use of
byproduct, source, or special nuclear material without restriction to
amount or atomic number, for sample analysis or instrument calibration
or associated with radioactive apparatus or components.
Date of issuance: April 17, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 39.
Facility Operating License No. DPR-7: This amendment revises the
license.
Date of initial notice in Federal Register: February 13, 2007 (72
FR 6788).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 17, 2007.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: December 29, 2006.
Brief description of amendments: The amendments revised Technical
Specification (TS) 3.4.1, ``Reactor Coolant System (RCS) Pressure,
Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,''
and TS 5.6.5, ``Core Operating Limits Report (COLR).'' This amendment
relocated the RCS DNB parameters for pressurizer pressure and RCS
average temperature to the COLR. In addition, TS 5.6.5 was revised to
add topical reports WCAP-8567-P-A, ``Improved Thermal Design
Procedure,'' and WCAP-11596-P-A, ``Qualification of the PHOENIX-P/ANC
Nuclear Design System for Pressurized Water Reactor Cores.''
Date of issuance: April 17, 2007.
Effective date: As of its date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: Unit 1--195; Unit 2--196.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: February 13, 2007 (72
FR 6786).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 17, 2007.
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: May 1, 2006, as supplemented
October 9, 2006, and February 21, 2007.
Brief description of amendments: The amendments relocate the main
steamline discharge radiation monitors (R46) from Technical
Specification (TS) 3/4.3.3.1, ``Radiation Monitoring Instrumentation''
to TS 3/4.3.3.7, ``Accident Monitoring Instrumentation.'' In addition,
the amendments modify TS definition 1.31, ``Source Check.''
Date of issuance: April 19, 2007.
Effective date: As of the date of issuance, to be implemented
within 30 days.
Amendment Nos.: 280 and 263.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs and the License.
Date of initial notice in Federal Register: July 18, 2006 (71 FR
40753). The supplements dated October 9, 2006, and February 21, 2007,
provided additional information that clarified the application, did not
expand the scope of the application as originally noticed, and did not
change the NRC staff's original proposed no significant hazards
consideration determination as published in the Federal Register on
July 18, 2006 (71 FR 40753).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 19, 2007.
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: June 7, 2006.
Brief description of amendments: The amendments delete the
Technical Specification (TS) requirements related to hydrogen
recombiners and hydrogen analyzers. The changes support the
implementation of a revision to Title 10 of the Code of Federal
Regulations, Section 50.44, ``Combustible gas control for nuclear power
reactors'' that became effective on October 16, 2003. A notice of
availability for this TS improvement using the consolidated line item
improvement process was published in the Federal Register on September
25, 2003 (68 FR 55416).
Date of issuance: April 19, 2007.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment Nos.: 281 and 264.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs and the License.
Date of initial notice in Federal Register: August 29, 2006 (71 FR
51231).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 19, 2007.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: December 21, 2006.
Description of amendment request: The amendments revised Technical
Specification (TS) Limiting Condition for Operation 3.10.1, and the
associated Bases, to expand its scope to include provisions for
temperature excursions greater than 212 [deg]F as a consequence of
inservice leak and hydrostatic testing, and as a consequence of scram
time testing initiated in conjunction with inservice leak or
hydrostatic testing, while considering operational conditions to be in
Mode 4.
Date of issuance: April 16, 2007.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment Nos.: 270, 299 & 258.
[[Page 26182]]
Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the TSs.
Date of initial notice in Federal Register: February 13, 2007 (72
FR 6791).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 16, 2007.
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: December 15, 2006.
Brief description of amendment: The amendment revised the Technical
Specifications to adopt NRC-approved Revision 4 to Technical
Specification Task Force (TSTF) Standard Technical Specification Change
Traveler TSTF-372, ``Addition of LCO [Limiting Condition for Operation]
3.0.8, Inoperability of Snubbers.'' The amendment added (1) a new LCO
3.0.8 addressing situations where one or more required snubbers are
unable to perform their associated support function(s) and (2) a
reference to LCO 3.0.8 in LCO 3.0.1, which describes when LCOs shall be
met.
Date of issuance: April 17, 2007.
Effective date: As of its date of issuance and shall be implemented
within 90 days of the date of issuance.
Amendment No.: 173.
Facility Operating License No. NPF-42: The amendment revised the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 3, 2007 (72 FR
154).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 17, 2007.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room (PDR), located at One White Flint North, Public File Area 01F21,
11555 Rockville Pike (first floor), Rockville, Maryland. Publicly
available records will be accessible from the Agencywide Documents
Access and Management System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2.
Interested persons should consult a current copy of 10 CFR 2.309, which
is available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland, and electronically on the Internet at the
[[Page 26183]]
NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If
there are problems in accessing the document, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected].
If a request for a hearing or petition for leave to intervene is filed
by the above date, the Commission or a presiding officer designated by
the Commission or by the Chief Administrative Judge of the Atomic
Safety and Licensing Board Panel, will rule on the request and/or
petition; and the Secretary or the Chief Administrative Judge of the
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------
\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protection order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer or the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear
Plant, Unit 3, Limestone County, Alabama
Date of application for amendment: April 6, 2007 (TS-460-T).
Brief description of amendment: This amendment approves a one-time
extension of the Completion Time for emergency diesel generator (EDG)
`3D' from 7 days to 14 days. The extension allows continued operation
while repairs, post-maintenance testing, and surveillance testing of
the subject EDG are completed.
Date of issuance: April 6, 2007.
Effective date: April 6, 2007, to be implemented within 30 days.
Amendment No.: 257.
Renewed Facility Operating License No. DPR-68: Amendment revises
the Technical Specifications.
The Commission's related evaluation of the amendment, finding of
emergency circumstances, and final determination of no significant
hazards consideration, are contained in a Safety Evaluation dated April
6, 2007.
Public comments requested as to proposed no significant hazards
consideration: No.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
[[Page 26184]]
NRC Section Chief: Thomas H. Boyce.
Dated at Rockville, Maryland, this 1st day of May 2007.
For the Nuclear Regulatory Commission.
Harold K. Chernoff,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. E7-8679 Filed 5-7-07; 8:45 am]
BILLING CODE 7590-01-P