[Federal Register Volume 70, Number 127 (Tuesday, July 5, 2005)]
[Notices]
[Pages 38712-38729]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-12987]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Application 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 June 10, 2005 to June 23, 2005. The last
biweekly notice was published on June 21, 2005 (70 FR 35735).
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.
[[Page 38713]]
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
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
[[Page 38714]]
the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC,
Attention: Rulemakings and Adjudications Staff at (301) 415-1101,
verification number is (301) 415-1966. A copy of the request for
hearing and petition for leave to intervene should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and it is requested that copies be
transmitted either by means of facsimile transmission to (301) 415-3725
or by e-mail to [email protected]. A copy of the request for
hearing and petition for leave to intervene should also be sent to the
attorney for the licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(I)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station (PVNGS),
Units 1, 2, and 3, Maricopa County, Arizona
Date of amendments request: May 26, 2005.
Description of amendments request: The amendments would revise the
Technical Specification (TS) requirements related to steam generator
(SG) tube integrity, consistent with those in NRC-approved Revision 4
to Technical Specification Task Force (TSTF) Standard Technical
Specification Change Traveler, TSTF-449, ``Steam Generator Tube
Integrity.'' The proposed amendment also includes changes to the
revised SG program in TS Section 5.5.9 to specify the SG tube
inspection length through the SG tubesheet and establish plugging
criteria in the inspected tubesheet region for the remaining original
SGs containing Alloy 600 mill annealed (MA) tubes. This change is being
proposed to establish conformance with the NRC position identified in
Generic Letter (GL) 2004-01, ``Requirements for Steam Generator Tube
Inspections.''
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 analysis that established the inspection length through the
SG tube sheet for the PVNGS Alloy 600 MA-tube SGs took into account
the reinforcing effect the tubesheet has on the external surface of
an expanded SG tube. Tube-bundle integrity will not be adversely
affected by the implementation of the revised tube inspection scope.
SG tube burst or collapse cannot occur within the confines of the
tubesheet; therefore, the tube burst and collapse criteria of draft
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR Steam
Generator Tubes,'' are inherently met. Any degradation below the
inspection length is shown by analyses and test results to be
acceptable, thereby precluding an event with consequences similar to
a postulated tube rupture event.
Tube burst is precluded for cracks within the tubesheet by the
constraint provided by the tubesheet. Thus, structural integrity is
maintained by the tubesheet constraint. However, a 360-degree
circumferential crack or many axially oriented cracks could permit
severing of the tube and tube pullout from the tubesheet under the
axial forces on the tube from primary to secondary pressure
differentials. Analysis and testing was performed to define the
length of non-degraded tubing that is sufficient to compensate for
the axial forces on the tube and thus prevent pullout. That length
is bounded by the inspection length proposed in this change.
In conclusion, incorporation of the revised inspection scope
into PVNGS TS maintains existing design limits and therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated.
Response: No.
The proposed performance based requirements are an improvement
over the requirements imposed by the current TS.
Implementation of the proposed Steam Generator Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the Steam Generator Program will be
an enhancement of SG tube performance. Primary to secondary leakage
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Tube-bundle integrity is expected to be maintained during all
plant conditions upon implementation of the proposed tube inspection
scope. Use of this scope does not introduce a new mechanism that
would result in a different kind of accident from those previously
analyzed. Even with the limiting circumstances of a complete
circumferential separation of a tube occurring below the inspection
length into the tubesheet, SG tube pullout is precluded and leakage
is predicted to be maintained within the Updated Final Safety
Analysis Report limits during all plant conditions.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety.
Response: No.
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
also isolate the radioactive fission products in the primary coolant
from the secondary system. In summary, the safety function of a SG
is maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the tube
integrity by implementing the Steam Generator Program to manage SG
tube inspection, assessment, repair, and plugging. The requirements
established by the Steam Generator Program are consistent with those
in the applicable design codes and standards and are an improvement
over the requirements in the current TS.
Upon implementation of the revised inspection scope, operation
with potential cracking below the Inspection Extent length in the
expansion region of the SG tubing will meet the margin of safety as
defined by Regulatory Guide (RG) 1.83 [Inservice Inspection of
Pressurized Water Reactor Steam Generator Tubes], draft RG 1.121
[Bases for Plugging Degraded PWRSteam Generator Tubes], and the
requirements of General Design Criteria 14, 15, 31, and 32 of
Appendix A to 10 CFR 50.
[[Page 38715]]
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix,
Arizona 85072-2034. NRC Acting Section Chief: Daniel S. Collins.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: June 3, 2005.
Description of amendments request: The proposed amendments would
revise the Updated Final Safety Analysis Report (UFSAR) for Palo Verde
Nuclear Generating Station (PVNGS), Units 1, 2 and 3. The proposed
amendments would reflect a modification performed by the licensee that
replaced the automatic water makeup function for the emergency diesel
generator jacket water cooling system with that of manual operator
actions.
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 emergency diesel generator (EDG) is a system that must
function in response to an accident that has been evaluated in
either Chapter 6 or 15 of the PVNGS UFSAR. It is designed to respond
to certain described accident scenarios. None of the accidents
evaluated are initiated within the EDG system. Therefore, this
request to allow the replacement of the automatic makeup feature(s)
with a manual feature can not increase the probability of an
accident previously postulated in the UFSAR.
None of the accidents evaluated which credit operation of the
EDG system require automatic fill of the DGCWS [Diesel Generator
Cooling Water System] in order to mitigate the consequences of the
accident. The fill system, whether automatic in nature as originally
designed or manual, simply maintains the EDG in the ready state.
Therefore, the proposed change does not involve a significant
increase in the 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 EDG is a piece of equipment important to safety. This
modification replaces the automatic water makeup function for the
EDG jacket water cooling system with that of manual operator
actions. The jacket water makeup is needed for normal leakage and
possible evaporation. Area walkdowns occur twice daily when the
diesel generator is in a standby mode (not running) and more
frequently (thirty minutes after initial loading and every two hours
while loaded) when the EDG is being tested or has responded to an
emergency event. The area operator walkdown procedures instruct the
operators to log the standpipe level and ensure it is in the normal
operating range. If the level is not, operators are required to
restore level and conduct further investigation of the condition and
notify appropriate personnel. This ensures that enough water remains
in the jacket water system to allow the diesel to remain operational
and evaluations are performed in order to detect any abnormal
leakrates. Therefore, the normal area operator walkdowns and
frequencies are adequate to ensure that sufficient jacket water
standpipe inventory is maintained.
With this modification, the EDG is still maintained and
monitored for proper conditions in a standby status to ensure that
it will respond to emergencies when called upon. Once the EDG
responds to an emergency signal and is loaded, its jacket water
system is required to be monitored every two hours to help ensure
that all parameters are observed and maintained for proper
operation, including its jacket water standpipe level.
So, with these measures in place it can be expected that the EDG
will be maintained capable of performing as designed to any
emergency safety signal. The [E]DG safety system and its support
jacket water cooling system do not initiate any accident events.
Therefore, the modification of this non-safety support system
cannot 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 PVNGS UFSAR states that the design basis function for the
emergency diesel generators is to provide a standby source of onsite
Class 1E AC power for the two trains of engineered safety features
equipment for safe plant shutdown and decay heat removal in the
event of loss of preferred (off-site) power. Supporting this design
basis function of supplying emergency power is the function of the
emergency diesel generator jacket cooling water system, which is to
remove rejected heat from each diesel engine at the rated design
load of the emergency diesel generator. The UFSAR further describes
the emergency diesel generator jacket cooling water surge tank
(standpipe), stating that the surge tank is sized to provide an
adequate reservoir to compensate for any minor leaks. The UFSAR also
described makeup to the jacket cooling water system as being
automatically actuated and provided from the safety-grade condensate
transfer system or manually from the demineralized water systems.
The subject modification replaced the automatic features with manual
operator action--the sources of the makeup water have not changed.
The PVNGS engineering analyses and the safety analyses that
demonstrate the functional goals and the design basis of the
emergency diesel generator system do not credit any makeup water
supply to the jacket cooling water system of the emergency diesel
generator for an initial 25 hours into an event. Operator monitoring
and manual makeup provides adequate control for maintaining the
DGCWS standpipe level, both for standby and loaded conditions. An
automatically actuated makeup water supply is not essential to the
safe and continued operation of the emergency diesel generator.
Makeup water is provided as a convenient source of water to
compensate for anticipated normal system losses and evaporation. It
is not provided to serve as an emergency source of makeup water to
the jacket cooling water system in the event of a major failure or
leak occurring within the jacket cooling water system.
Makeup to the system is required to compensate for normal
expected system losses, minor leaks, and evaporation. In addition,
an engineering calculation has been performed to address 10 CFR 50,
Appendix R concerns, which demonstrates that no operator action is
required or credited during the first twenty-five hours of emergency
diesel generator loaded operation provided that the initial water
level is at the specified minimum level. This twenty-five hour
period before operator intervention, which is assumed to occur,
sufficiently bounds the thirty minutes of no operator action that is
normally assumed in most of the accident analyses.
In addition, the area operator walkdown procedures instruct the
operators to log the standpipe level and ensure it is in the normal
operating range. If the level is not, operators are required to
restore level and conduct further investigation of the condition and
notify appropriate personnel. This ensures that enough water remains
in the jacket water system to allow the diesel to remain operational
and evaluations are performed in order to detect any abnormal
leakrates.
Therefore, APS has concluded that the proposed license amendment
request does not involve a significant reduction in a margin of
safety.
Based on the above, Arizona Public Service Company (APS)
concludes that the proposed amendment presents no significant
hazards consideration under the standards set forth in 10 CFR
50.92(c), and, accordingly, a finding of ``no significant hazards
consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
[[Page 38716]]
Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix,
Arizona 85072-2034.
NRC Section Chief: Daniel S. Collins.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: June 7, 2005.
Description of amendments request: The proposed amendment would
revise Technical Specification (TS) 3.1.1, ``Shutdown Margin,'' to
modify Required Action B.1 restricting a positive reactivity addition.
The proposed amendment would also correct an administrative error
regarding an incorrect TS reference in TS 3.4.17, ``Special Test
Exception RCS [reactor coolant system] Loops--Modes 4 and 5.''
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The intent of this change is to clarify a Technical
Specification involving positive reactivity additions to the
shutdown reactor so that small, controlled, safe insertions of
positive reactivity will be allowed where they are now categorically
prohibited, posing a potential conflict between two required
actions. These controlled activities could result in a slight change
in the probability of an event occurring as a RCS manipulation that
is currently prohibited would now be allowed. However, RCS
manipulations are rigidly controlled to minimize the possibility of
a significant reactivity increase.
In addition, there is sufficient shutdown margin available in
this condition to allow for slight reactivity changes without
significantly increasing the probability of an accident previously
evaluated.
The proposed change involving positive reactivity additions does
not permit the shutdown margin required by the Technical
Specifications to be reduced. While the proposed change will permit
changes in the discretionary boron concentration above the Technical
Specification requirements, this excess concentration is not
credited in the Updated Final Safety Analysis Report safety
analysis. Because the initial conditions assumed in the safety
analysis are preserved, no increase in the consequence of an
accident previously evaluated would occur. These small changes are
within the required shutdown margin, therefore, there is no increase
in the consequence of an accident previously evaluated.
The administrative error was in the marked up Technical
Specification pages submitted with a proposed change. The correct
Technical Specification number was provided in the proposal letter
and was used by the staff in the discussion for accepting the
proposed change. Correcting this administrative error does not
change the significant hazards discussion previously submitted.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different kind
of accident from any accident previously evaluated.
This proposed change involving positive reactivity addition
allows for a minor plant operational adjustment without adversely
impacting the safety analysis required shutdown margin. It does not
involve any change to plant equipment or the shutdown margin
requirements in the Technical Specifications.
The administrative error was in the marked up Technical
Specification pages submitted with a proposed change. The correct
Technical Specification number was provided in the proposal letter
and was used by the staff in the discussion for accepting the
proposed change. Correcting this administrative error does not
change the significant hazards discussion previously submitted.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in [a] margin of
safety.
The margin of safety in Modes 3, 4 and 5 is preserved by the
calculated shutdown margin which prevents an inadvertent
criticality. The proposed change involving positive reactivity
addition will permit reductions in discretionary shutdown margin
that is beyond Technical Specification requirements. However, the
shutdown margin required by the Technical Specifications is not
changed. By not impacting the shutdown margin, the margin of safety
is not affected.
The administrative error was in the marked up Technical
Specification pages submitted with a proposed change. The correct
Technical Specification number was provided in the proposal letter
and was used by the staff in the discussion for accepting the
proposed change. Correcting this administrative error does not
change the significant hazards discussion previously submitted.
Therefore, the proposed change will 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
amendments request involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Generation Group, LLC, 750 East Pratt Street,
17th floor, Baltimore, MD 21202.
NRC Section Chief: Richard J. Laufer.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: June 7, 2005.
Description of amendments request: The proposed amendment would
revise the Technical Specifications (TSs) to eliminate the use of the
defined term Core Alterations. The proposed amendment would incorporate
the changes reflected in TS Task Force (TSTF) Travelers 471-T (TSTF-
471-T) and TSTF-51-A. In addition, the proposed amendment would revise
TS 3.9.2, ``Nuclear Instrumentation,'' by replacing ``Core
Alterations'' with ``positive reactivity additions'' in the required
action for an inoperable source range monitor during refueling
operations. The limiting conditions for operation in TS 3.9.4,
``Shutdown Cooling (SDC) and Coolant Recirculation--High Water Level,''
would also be revised by replacing ``core alterations'' with ``movement
of fuel assemblies within containment.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change eliminates the use of the defined term CORE
ALTERATIONS from the Technical Specifications. Core alterations are
not an initiator of any accident previously evaluated except a fuel
handling accident. Those revised Technical Specifications that
protect the initial conditions of a fuel handling accident also
require the suspension of movement of irradiated fuel assemblies,
which protects the initial condition of a fuel handling accident.
Therefore, suspension of CORE ALTERATIONS do not affect the
initiators of the accidents previously evaluated and suspension of
CORE ALTERATIONS does not affect the mitigation of the accidents
previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
No new or different accidents result from utilizing the proposed
change. The changes
[[Page 38717]]
do not involve a physical modification of the plant (i.e., no new or
different type of equipment will be installed) or a significant
change in the methods governing normal plant operation. In addition,
the changes do not impose any new or different requirements. The
changes do not alter assumptions made in the safety analysis. The
proposed changes are consistent with the safety analysis
assumptions.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
Only two accidents are postulated to occur during plant
conditions where CORE ALTERATIONS may be made: A fuel handling
accident and a boron dilution accident. Suspending movement of
irradiated fuel assemblies prevents a fuel handling accident. Also
requiring the suspension of CORE ALTERATIONS is redundant to
suspending movement of irradiated fuel assemblies and does not
increase the margin of safety. CORE ALTERATIONS have no effect on a
boron dilution accident. Core components are not involved in the
initiation or mitigation of a boron dilution accident. Therefore,
CORE ALTERATIONS have no effect on the margin of safety related to a
boron dilution accident.
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
amendments request involves no significant hazards consideration.
Attorney for licensee: Carey Fleming, Sr. Counsel--Nuclear
Generation, Constellation Generation Group, LLC, 750 East Pratt Street,
17th floor, Baltimore, MD 21202.
NRC Section Chief: Richard J. Laufer.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina and Docket
Nos. 50-369 and 50-370, McGuire Nuclear Station, Units 1 and 2,
Mecklenburg County, North Carolina
Date of amendment request: October 27, 2004.
Description of amendment request: The amendments would revise the
facility operating licenses (FOLs) to remove a license condition that
limits the maximum rod average burnup for any rod to 60 GWd/mtU. This
deletion would allow the 62 GWd/mtU limit, approved by the NRC, as
documented in Duke Topical Report DPC-2009-P-A, to become the burnup
limit. The amendments would also revise both of the station's Updated
Final Safety Analysis Reports (Section 4.0) to include a new discussion
of the fuel burnup limit. Additionally, approval would allow Duke to
make an administrative revision to Duke Topical Report DPC-NE-2009-P-A,
Revision 2, to reference the approval of these amendments and to
reflect removal of the current license condition. Furthermore, the
amendments would remove the McGuire FOL Section 2.E, that lists
reporting requirements with regard to Maximum Power Level, Fire
Protection, Protection of the Environment (Unit 2 FOL only), and
Physical Protection. It would also remove the Catawba FOL Section 2.F,
that lists reporting requirements with regard to Maximum Power Level,
Updated Final Safety Analysis Report, Antitrust Conditions, Fire
Protection, and Additional Conditions.
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. Would implementation of the changes proposed in this LAR
[License Amendment Request] involve a significant increase in the
probability or consequences of an accident previously evaluated?
No, deletion of the fuel burnup limit currently stated as an
additional license condition in the McGuire and Catawba Facility
Operating Licenses has no impact on accident probabilities. Further,
as determined in the NRC's environmental assessment which supports
the increased burnup limit (NUREG/CR-6703, Environmental Effects of
Extending Fuel Burnup Above 60 GWd/mtU), the potential environmental
consequences of postulated accidents are not expected to increase
significantly with increased burnup. Duke concurs with this
assessment conclusion for the burnup range in this LAR.
The deletion of the reporting requirements from the FOLs is
solely administrative. No plant equipment or accident analyses will
be affected by this deletion.
2. Would implementation of the changes proposed in this LAR
create the possibility of a new or different kind of accident from
any accident previously evaluated?
No, implementation of this amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated. No new accident causal mechanisms will be
created as a result of the NRC approval of this LAR. No changes are
being made to the plant which will introduce any new accident causal
mechanisms. This amendment does not otherwise impact any plant
structures, systems, or components that are accident initiators;
therefore, no new accident types are being created.
3. Would implementation of the changes proposed in this LAR
involve a significant reduction in a margin of safety?
No, margin of safety is related to the confidence in the ability
of the fission product barriers to perform their design functions
during and following an accident situation. These barriers include
the fuel cladding, the reactor coolant system, and the containment
system. These barriers are not significantly affected by the changes
proposed in this LAR. The effect of the increased burnup on fuel
cladding was considered in the NRC's environmental assessment
supporting the increase in the fuel burnup limit. Further, the
proposed limit is equal to that approved for the fuel rod cladding
at McGuire and Catawba.
The deletion of the reporting requirements from the FOLs is
solely administrative in nature. No plant equipment or accident
analyses will be affected by this deletion.
The margin of safety is established through the design of the
plant structures, systems, components, the parameters within which
the plant is operated, and the establishment of the setpoints for
the actuation of equipment relied upon to respond to an event, and
thereby protect the fission product barriers. The proposed changes
have no significant impact on any of these considerations in regard
to the physical plant or the manner in which it is operated.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina 28201-1006.
NRC Section Chief: Evangelos C. Marinos
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: October 11, 2004.
Description of amendment request: The proposed amendments apply to
Technical Specifications 3.8.1, ``AC Sources--Operating,'' and 3.8.9,
``Distribution Systems--Operating.'' They would extend several
completion times and would modify several Surveillance Requirement (SR)
Notes. Additionally, they would correct a recently identified non-
conservative situation that currently exists with SR 3.8.1.4.
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
[[Page 38718]]
consideration, which is presented below:
First Standard
Will implementation of the changes proposed in this license
amendment request involve a significant increase in the probability
or consequences of an accident previously evaluated?
No. The changes proposed in this license amendment request
increase the Technical Specifications Completion Times for the
emergency diesel generators and electrical power and distribution
systems. Increasing these Completion Times will not cause a
significant increase in the probability or consequences of an
accident which has been previously evaluated. This license amendment
request is supported by an extensive risk-informed study performed
by the nuclear industry and documented in a topical report and
Technical Specifications Task Force travelers that have been
submitted for NRC review and approval. Within this study, the risk
impacts of increasing the Completion Times were calculated and
compared against the acceptability guidelines contained in the
applicable regulatory guides and found to be acceptable. The
emergency diesel generators and electrical power and distribution
systems and equipment affected by this license amendment request
will remain highly reliable. Thus there will be no significant
increase in the probability or consequences of an accident which has
been previously evaluated.
The proposed changes that modify Surveillance Requirement notes
are consistent with an NRC [Nuclear Regulatory Commission]-approved
industry initiative. Implementation of these changes will require
that the plant's risk be managed. Thus there will be no significant
increase in the probability or consequences of an accident which has
been previously evaluated.
The proposed change that corrects the non-conservative
Surveillance Requirement only increases a Technical Specifications
parameter value in the conservative direction. Thus this change will
not contribute to any increase in the probability or consequences of
an accident which has been previously evaluated.
Second Standard
Will implementation of the changes proposed in this license
amendment request create the possibility of a new or different kind
of accident from any accident previously evaluated?
No. The proposed changes would create no new accidents since no
changes are being made that introduce any new accident casual
mechanisms. The deterministic evaluation that supports this license
amendment request consisted of a review of plant systems and safety
functions impacted by entry into the expanded Completion Times, the
performance of testing in previously prohibited operating modes, or
increasing a Technical Specification mandated parameter in the
conservative direction. The emergency diesel generators and
electrical power and distribution systems were quantitatively and
qualitatively assessed. It was determined that no new accidents or
transients would be introduced by the proposed changes.
Third Standard
Will implementation of the changes proposed in this license
amendment request involve a significant reduction in a margin of
safety?
No. The impact of the proposed changes on the safety margins was
considered in the deterministic evaluations that support this
license amendment request. Extending the Completion Times,
performing testing activities to confirm operability, or
conservatively increasing a Technical Specification controlled
parameter does not adversely impact any assumptions or inputs in the
transient analyses contained in the McGuire Updated Final Safety
Analysis Report (UFSAR). The proposed changes have no negative
impact upon the ability of the fission product barriers (fuel
cladding, the reactor coolant system, and the containment system) to
perform their design functions during and following an accident
situation. Additionally, the proposed changes have no adverse impact
on setpoints or limits established or assumed within the UFSAR.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c))
are satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Section Chief: Evangelos C. Marinos.
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: May 24, 2005.
Description of amendment request: The proposed amendment would
revise the steam generator (SG) tube inspection scope for Byron
Station, Unit 2 for Refueling Outage 12 and the subsequent operating
cycle. The proposed changes modify the inspection requirements for
portions of SG tubes within the hot leg tubesheet region of the SGs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The previously analyzed accidents are initiated by the failure
of plant structures, systems, or components. The proposed changes
that alter the SG inspection criteria do not have a detrimental
impact on the integrity of any plant structure, system, or component
that initiates an analyzed event. The proposed changes will not
alter the operation of, or otherwise increase the failure
probability of any plant equipment that initiates an analyzed
accident. Therefore, the proposed change does not involve a
significant increase in the probability of an accident previously
evaluated.
Of the applicable accidents previously evaluated, the limiting
transients with consideration to the proposed changes to the SG tube
inspection criteria, are the SG tube rupture (SGTR) event and the
steam line break (SLB) accident.
During the SGTR event, the required structural integrity margins
of the SG tubes will be maintained by the presence of the SG
tubesheet. SG tubes are hydraulically expanded in the tubesheet
area. Tube rupture in tubes with cracks in the tubesheet is
precluded by the constraint provided by the tubesheet. This
constraint results from the hydraulic expansion process, thermal
expansion mismatch between the tube and tubesheet and from the
differential pressure between the primary and secondary side. Based
on this design, the structural margins against burst, discussed in
Regulatory Guide (RG) 1.121, ``Bases for Plugging Degraded PWR
[pressurized water reactor] SG Tubes,'' are maintained for both
normal and postulated accident conditions.
The proposed changes do not affect other systems, structures,
components or operational features. Therefore, the proposed changes
result in no significant increase in the probability of the
occurrence of a SGTR accident.
At normal operating pressures, leakage from primary water stress
corrosion cracking (PWSCC) below the proposed limited inspection
depth is limited by both the tube-to-tubesheet crevice and the
limited crack opening permitted by the tubesheet constraint.
Consequently, negligible normal operating leakage is expected from
cracks within the tubesheet region. The consequences of an SGTR
event are affected by the primary-to-secondary leakage flow during
the event. Primary-to-secondary leakage flow through a postulated
broken tube is not affected by the proposed change since the
tubesheet enhances the tube integrity in the region of the hydraulic
expansion by precluding tube deformation beyond its initial
hydraulically expanded outside diameter.
The probability of a SLB is unaffected by the potential failure
of a SG tube as this failure is not an initiator for a SLB.
The consequences of a SLB are also not significantly affected by
the proposed changes. During a SLB accident, the reduction in
pressure above the tubesheet on the shell side of the SG creates an
axially uniformly distributed load on the tubesheet due to the
reactor coolant system pressure on the underside of the tubesheet.
The resulting bending action constrains the tubes in the tubesheet
thereby restricting primary-to-secondary leakage below the midplane.
Primary-to-secondary leakage from tube degradation in the
tubesheet area during the
[[Page 38719]]
limiting accident (i.e., SLB) is limited by flow restrictions
resulting from the crack and tube-to-tubesheet contact pressures
that provide a restricted leakage path above the indications and
also limit the degree of potential crack face opening as compared to
free span indications. The primary-to-secondary leak rate during
postulated SLB accident conditions would be expected to be less than
that during normal operation for indications near the bottom of the
tubesheet (i.e., including indications in the tube end welds). This
conclusion is based on the observation that while the driving
pressure causing leakage increases by approximately a factor of two,
the flow resistance associated with an increase in the tube-to-
tubesheet contact pressure, during a SLB, increases by up to
approximately a factor of three. While such a leakage decrease is
logically expected, the postulated accident leak rate could be
conservatively bounded by twice the normal operating leak rate if
the increase in contact pressure is ignored. Since normal operating
leakage is limited to less than 0.104 gpm (150 gpd) per TS 3.4.13,
``RCS Operational Leakage,'' the associated accident condition leak
rate, assuming all leakage to be from lower tubesheet indications,
would be bounded by approximately 0.2 gpm. This value is well within
the assumed accident leakage rate of 0.5 gpm discussed in Updated
Final Safety Analysis Table 15.1-3, ``Parameters Used in Steam Line
Break Analyses.'' Hence it is reasonable to omit any consideration
of inspection of the tube, tube end weld, bulges/overexpansions or
other anomalies below 17 inches from the top of the hot leg
tubesheet. Therefore, the consequences of a SLB accident remain
unaffected.
Based on the above discussion, the proposed changes do not
involve an increase in the consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not involve the use or installation of
new equipment and the currently installed equipment will not be
operated in a new or different manner. No new or different system
interactions are created and no new processes are introduced. The
proposed changes will 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 significant reduction in a
margin of safety?
Response: No.
The proposed changes maintain the required structural margins of
the SG tubes for both normal and accident conditions. Nuclear Energy
Institute (NEI) 97-06, ``Steam Generator Program Guidelines,''
Revision 1 and Regulatory Guide (RG) 1.121, ``Bases for Plugging
Degraded PWR Steam Generator Tubes,'' are used as the bases in the
development of the limited hot leg tubesheet inspection depth
methodology for determining that SG tube integrity considerations
are maintained within acceptable limits. RG 1.121 describes a method
acceptable to the NRC for meeting General Design Criteria (GDC) 14,
``Reactor coolant pressure boundary,'' GDC 15, ``Reactor coolant
system design,'' GDC 31, ``Fracture prevention of reactor coolant
pressure boundary,'' and GDC 32, ``Inspection of reactor coolant
pressure boundary,'' by reducing the probability and consequences of
a SGTR. RG 1.121 concludes that by determining the limiting safe
conditions for tube wall degradation the probability and
consequences of a SGTR are reduced. This RG uses safety factors on
loads for tube burst that are consistent with the requirements of
Section III of the American Society of Mechanical Engineers (ASME)
Code.
For axially oriented cracking located within the tubesheet, tube
burst is precluded due to the presence of the tubesheet. For
circumferentially oriented cracking, Westinghouse letter LTR-CDME-
05-32-P, ``Limited Inspection of the Steam Generator Tube Portion
Within the Tubesheet at Byron Unit 2 and Braidwood Unit 2,''
Revision 1, dated May 2005, defines a length of degradation free
expanded tubing that provides the necessary resistance to tube
pullout due to the pressure induced forces, with applicable safety
factors applied. Application of the limited hot leg tubesheet
inspection depth criteria will preclude unacceptable primary-to-
secondary leakage during all plant conditions. The methodology for
determining leakage provides for large margins between calculated
and actual leakage values in the proposed limited hot leg tubesheet
inspection depth criteria.
Therefore, the proposed changes do not involve a significant
hazards consideration under the criteria set forth in 10 CFR
50.92(c).
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. Thomas S. O'Neill, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: Gene Y. Suh.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: May 11, 2005. The proposed amendment
supercedes, in its entirety, a previous amendment request dated April
29, 2004, published in the Federal Register on May 25, 2004 (69 FR
29766).
Description of amendment request: The proposed amendment would
revise technical specification (TS) 3/4.4.10, ``Reactor Coolant
System--Structural Integrity, ASME Code Class 1, 2, and 3 Components,''
to allow a one-time extension of the surveillance interval for the
reactor vessel internals vent valves from September 2005 to March 2006.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed one-time surveillance interval exception does not
alter the design, operation, or testing method of any structure,
system, or component. Therefore, the proposed change does not
involve a significant increase in the probability of an accident
previously evaluated. In addition, no accident initiators are
affected and no previously analyzed accident scenario is changed.
Initiating conditions and assumptions remain as previously analyzed.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed one-time surveillance interval exception does not
alter the design, operation, or testing method of any structure,
system, or component. The proposed change does not introduce any new
or different accident initiators. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed one-time surveillance interval exception does not
affect the capabilities of the Reactor Vessel Internals Vent Valves.
Therefore, the proposed change will not involve 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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Gene Y. Suh.
[[Page 38720]]
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: May 22, 2005.
Description of amendment request: The proposed amendment would
adopt a qualified alternate repair criteria (ARC) for axial tube end
cracking (TEC) indications in the Davis-Besse Nuclear Power Station,
Unit 1 once-through steam generator tubes. Specifically, the proposed
amendment would revise the technical specification surveillance
requirements for steam generator tube inservice inspection to include
the TEC ARC. The technical basis for the ARC is provided in Babcock &
Wilcox Owners Group Topical Report BAW-2346P, ``Alternate Repair
Criteria for Tube End Cracking in the Tube-to-Tubesheet Roll Joint of
Once-Through Steam Generators,'' dated April 1999.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment does not increase the probability of any
accident. Steam generator tube failure is an initiating condition
for the steam generator tube rupture (SGTR) accident. The proposed
TEC ARC does not affect the probability of an SGTR because the TEC
ARC is limited to crack indications that are precluded from burst
due to the presence of the tubesheet. Therefore, the proposed change
does not involve a significant increase in the probability of an
accident previously evaluated.
The proposed amendment does not increase the consequences of any
previously evaluated accident. Primary-to-secondary leakage affects
the radiological consequences of accidents evaluated in the Updated
Safety Analysis Report. The proposed amendment may result in an
increase in post-accident primary-to-secondary leakage. Analyses
have been performed to determine the expected post-accident leakage
from each TEC left in service. The proposed amendment would impose
inservice inspection and leakage assessment requirements that would
ensure that the expected post-accident primary-to-secondary leakage
through TECs and all other sources is maintained below the value
assumed in the accident analyses. Therefore, the proposed change
does not involve a significant increase in the consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed TEC ARC does not introduce any new failure modes or
accident scenarios. Analyses have demonstrated that structural and
leakage integrity is maintained for normal operating and accident
conditions. Any failure of a tube from a TEC would be bounded by the
SGTR analysis. 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 amendment does not reduce the structural margin of
the steam generator tubes. Structural integrity of the tube is
maintained since the TEC ARC is limited to crack indications that
are precluded from burst due to the presence of the tubesheet. The
proposed amendment would impose inservice inspection and leakage
assessment requirements that will ensure that the expected post-
accident primary-to-secondary leakage through TECs and all other
sources is maintained below the value assumed in the accident
analyses. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308
NRC Section Chief: Gene Y. Suh.
Florida Power and Light Company, Docket Nos. 50-335 and 50-389, St.
Lucie Nuclear Plant, Units 1 and 2, St. Lucie County, Florida
Date of amendment request: April 21, 2005.
Description of amendment request: The submittal requests revision
to several Technical Specifications (TSs) using seven TS Task Force
(TSTF) generic changes. The seven TSTFs (nos. 5, 65, 101, 258, 299,
308, and 361) delete redundant safety limit violation notification
requirements; adopt use of generic titles for utility positions; change
the auxiliary feedwater pump test frequency to be consistent with the
inservice test program frequency; remove redundant requirements and add
other requirements to Section 5.0, Administrative Controls; clarify the
meaning of ``refueling cycle'' for system integrated leak test
intervals in the Primary Coolant Sources Outside Containment program;
clarify the requirements regarding the frequency of testing for
cumulative and projected dose contributions from radioactive effluents;
and add a note to the residual heat removal requirements during Mode 6
low water level operations that allows one required residual heat
removal (RHR) loop to be inoperable for up to 2 hours for surveillance
testing provided the other RHR loop is operable and in operation. In
addition, the proposed amendments revise the TSs to adopt the Improved
Standard Technical Specification (ISTS) requirements for remote
shutdown instrumentation and the ISTS actions and action times for
accident monitoring instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed changes revise administrative requirements,
actions, action times, surveillance requirements, and surveillance
frequencies. The revised requirements are not an initiator of any
accident previously evaluated. As a result, the probability of any
accident previously evaluated is not significantly increased by the
proposed changes. The Technical Specifications continue to require
the systems, structures, and components associated with the revised
requirements to be operable. Therefore, any mitigation functions
assumed in the accident analyses will continue to be performed. As a
result, the consequences of any accident previously evaluated are
not significantly increased. Therefore, the proposed amendments do
not involve a significant increase in the probability or
consequences of any accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any previously evaluated.
The proposed changes do not alter the design or physical
configuration of the plant. No changes are being made to the plant
that would introduce any new accident causal mechanisms. Therefore,
operation of the facility in accordance with the proposed amendments
do not create the possibility of a new or different kind of accident
from any previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The proposed changes do not change the design or function of
plant equipment. The proposed changes do not significantly reduce
the level of assurance that any associated plant equipment will be
available to perform its function. The proposed changes provide
[[Page 38721]]
operating flexibility without significantly affecting plant
operation. Therefore, operation of the facility in accordance with
the proposed amendments would not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Michael L. Marshall, Jr.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: May 25, 2005.
Description of amendment request: The proposed amendment would
delete from the Cooper Nuclear Station (CNS) Technical Specifications
(TSs) temporary notes that have expired and are no longer in effect.
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.
Deleting temporary notes that have expired from the CNS TS does
not impact the plant design or how the plant is operated, nor does
it affect any of the conditions that could cause an accident. Thus,
this change does not result in a significant increase in the
probability of an accident previously evaluated. Removing the
expired temporary notes does not reduce the requirements for
maintaining systems needed to mitigate postulated accidents as
described in the CNS Updated Safety Analysis Report. Thus, this
change does not result in a significant increase in the consequences
of an accident previously evaluated. Therefore, the proposed changes
do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Deleting temporary notes that have expired does not involve a
change to the plant design or to how the plant is operated.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
Deleting temporary notes that have expired does not result in a
relaxation of any limit associated with the performance of systems
required to mitigate postulated accidents, nor does it reduce any of
the requirements for maintaining those systems. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: David Terao.
R. E. Ginna Nuclear Power Plant, LLC, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: September 30, 2004, as supplemented on
May 28, 2005.
Description of amendment request: The proposed amendment would
revise the information in the Updated Final Safety Analysis Report
regarding the application of leak-before-break methodology to the
accumulator A and B lines and the pressurizer surge line. The
application of leak-before-break methodology would permit the exclusion
of these lines from the evaluation of dynamic effects associated with
postulated high energy line breaks.
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 previous[ly] evaluated?
Response: No.
The proposed changes use an approved fracture mechanics
methodology, in accordance with 10 CFR [Part] 50, Appendix A, GDC
[General Design Criterion] -4 to demonstrate that the probability of
fluid system rupture for these lines attached to the Reactor Coolant
System is extremely low under conditions associated with the design
basis for the piping.
The proposed changes do not adversely affect accident initiators
or precursors nor significantly alter the design assumptions,
conditions, and configuration of the facility or the manner in which
the plant is operated and maintained. The proposed changes do not
adversely 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. Further, the proposed changes do not increase
the types and amounts of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupation/public radiation exposures. The proposed changes do not
affect the probability of an accident occurring since they reflect a
change in plant design basis that is consistent with current
Regulations. The proposed changes cannot increase the consequences
of postulated accidents since LOCA [loss-of-coolant accident] and
methods containment analysis will not be changed. Therefore, the
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not create the possibility of a new or
different kind of accident, since it simply provides an analytical
justification for demonstrating that the probability of a fluid
system rupture is extremely small. Leak-before-break justifications
per GDC-4 still require that ECCS [emergency core cooling system],
containment, and EQ [environmental qualification] requirements be
maintained consistent with the original postulated accident
assumptions--only protection from dynamic effects is modified.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes apply very conservative approved analytical
methods to demonstrate that the probability of a fluid system
rupture is very low. This analysis justifies differences in
protection from dynamic effect [and] is associated with these
extremely low probability ruptures. For overall ECCS, containment,
and EQ requirements, there will be no changes to the licensing
basis.
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: Daniel F. Stenger, Ballard Spahr Andrews &
Ingersoll, LLP, 601 13th Street, NW., Suite 1000 South, Washington, DC
20005.
NRC Section Chief: Richard J. Laufer.
[[Page 38722]]
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: March 8, 2005.
Description of amendment request: The amendments proposed by
Southern Nuclear Operating Company (SNC) would revise the Technical
Specifications (TS) to delete Function 11, Reactor Coolant Pump (RCP)
Breaker Position, in TS 3.3.1, ``Reactor Trip System (RTS)
Instrumentation.''
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?
No. The proposed changes do not significantly increase the
probability or consequences of an accident previously evaluated in
the Updated Final Safety Analysis Report (UFSAR). All of the safety
analyses have been evaluated for impact. The elimination of RCP
Breaker Position reactor trip will not initiate any accident;
therefore, the probability of an accident has not been increased. An
evaluation of dose consequences, with respect to the proposed
changes, indicates there is no impact due to the proposed changes
and all acceptance criteria continue to be met. Therefore, these
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 previously evaluated?
No. The proposed changes do not create the possibility of a new
or different kind of accident than any accident already evaluated in
the UFSAR. No new accident scenarios, failure mechanisms or limiting
single failures are introduced as a result of the proposed changes.
The changes have no adverse effects on any safety-related system.
Therefore, all accident analyses criteria continue to be met and
these changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
No. The proposed changes do not involve a significant reduction
in a margin of safety. All analyses that credit the RCS Low Flow
reactor trip function have been reviewed and no changes to any
inputs are required. The evaluation demonstrated that all applicable
acceptance criteria are met. Therefore, the proposed changes do not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief: Evangelos C. Marinos.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: June 2, 2005.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) 3.4.6.1, ``Reactor Coolant System
Leakage Detection Systems,'' to specifically require only one
containment radioactivity monitor (particulate channel) to be operable
in Modes 1, 2, 3 and 4. Additionally, corresponding changes to the
Surveillance Requirement (SR) 4.4.6.1 and 4.4.6.2.1, ``Reactor Coolant
System Operational Leakage,'' are also requested.
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. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change has been evaluated and determined to not
increase the probability or consequences of an accident previously
evaluated. The proposed change does not make any hardware changes
and does not alter the configuration of any plant system, structure
or component (SSC). The proposed change only removes the containment
atmosphere gaseous radioactivity monitor as an option for meeting
the operability requirement for TS 3.4.6.1, and correspondingly from
the requirements of SR 4.4.6.1 and 4.4.6.2.1.a. Therefore, the
probability of occurrence of an accident is not increased. The TS
will continue to require diverse means of leakage detection
equipment, thus ensuring that leakage due to cracks would continue
to be identified prior to breakage and the plant shutdown
accordingly. Additionally, the proposed change is not modeled in the
South Texas Project probabilistic risk assessment and has no impact
on core damage frequency or large early release frequency.
Therefore, the consequences of an accident are not increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve the use or installation of
new equipment and the currently installed equipment will not be
operated in a new or different manner. No new or different system
interactions are created and no new processes are introduced. The
proposed changes will not introduce any new failure mechanisms,
malfunctions, or accident initiators not already considered in the
design and licensing bases. The proposed change does not affect any
SSC associated with an accident initiator. 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.
The proposed change does not make any alteration to any RCS
leakage detection components. The proposed change only removes the
gaseous channel of the containment atmosphere radioactivity monitor
as an option for meeting the operability requirement for TS 3.4.6.1,
and correspondingly from the requirements of SR 4.4.6.1 and
4.4.6.2.1.a. The proposed amendment continues to require diverse
means of leakage detection equipment with capability to promptly
detect RCS leakage. Although not required by TS, additional diverse
means of leakage detection capability are available. Based on this
evaluation, the proposed change does not involve a significant
reduction in a margin of safety.
Based upon the NRC staff's review, it appears that the standards of
10 CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the request for amendments involves no significant
hazards consideration.
Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Section Chief: David Terao.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: April 27, 2005.
Description of amendment request: The proposed amendment would
revise the applicability for Items 18.A and 18.B of Technical
Specification (TS) Table 3.3-1, ``Reactor Trip System
Instrumentation,'' and TS Table 4.3-1, ``Reactor Trip System
Instrumentation Surveillance Requirements.'' This change will add a
footnote that indicates that the Mode 1 applicability is limited to
operation above the P-9 (50-percent rated thermal power) value.
Additionally, the action for an
[[Page 38723]]
inoperable turbine stop valve closure channel is being revised to be
consistent with the design of this function. Finally, an option
consistent with the latest standard TSs (NUREG-1431, Revision 3) is
added to permit a reduction in thermal power to below the P-9 interlock
within 10 hours for an inoperable turbine stop valve closure channel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes revise the applicability and actions for
inoperable reactor trip functions from a turbine trip event. These
changes do not alter these functions physically or how they are
maintained. By clarifying the proper applicability and enhancing the
actions for these functions the availability of these trips and
compensatory measures for inoperable conditions are improved. The
availability change implements the required conditions for turbine
trip operability that are consistent with their ability to perform
the reactor trip functions. The action changes correct inappropriate
requirements for minimum channels to be operable and the allowance
to bypass channels in consideration of the logic design for the
turbine stop valve closure channels. The change to allow power
reduction as an alternative to tripping an inoperable channel for
the turbine stop valve closure channels, provides a more
conservative response than currently allowed.
Since these changes will not affect the ability of these trips
to perform the initiation of reactor trips when appropriate, the
offsite dose consequences for an accident will not be impacted.
Equally, the potential to cause an accident is not affected because
no plant system or component has been altered by the proposed
changes. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes only affect the applicability and action
requirements for the turbine trip functions. This does not affect
any physical features of the plant or the manner in which these
functions are utilized. The proposed applicability will require the
functions to be operable when they are able to perform their trip
functions. The actions will handle inoperable channels such that
their safety function will be satisfied or the unit will be placed
in a condition that does not require these trip functions.
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 changes do not alter any plant setpoints or
functions that are assumed to actuate in the event of postulated
accidents. In fact, the proposed changes do not alter any plant
feature and only alters the requirements for when the function must
be operable and the actions to take should a channel become
inoperable during these conditions. The proposed changes ensure the
functionality of the turbine trips when assumed in the analysis and
provides actions for inoperable channels that preserve the safety
functions for accident mitigation. 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee Date of
amendment request: April 27, 2005.
Description of amendment request: The proposed amendment would
relocate a number of technical specification (TS) requirements to the
Technical Requirements Manual (TRM).
The proposed amendment would relocate the provisions for TS 3.1.3.4
(Rod Drop Time), TS 3.3.2 (Movable Incore Detectors), TS 3.3.3.4
(Meteorological Instrumentation), TS 3.4.7 (Reactor Coolant System
Chemistry), TS 3.4.11 (Reactor Coolant System Head Vents), TS 3.7.2
(Steam Generator Pressure and Temperature Limitations), TS 3.7.10
(Sealed Source Contamination), TS 3.9.5 (Refueling Operations
Communications), and TS 3.9.6 (Manipulator Crane).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change only relocates requirements to TRM that are
not required to be included in the TSs in accordance with 10 CFR
50.36. Changes to the TRM require evaluations and reviews in
accordance with 10 CFR 50.59 to ensure that the health and safety of
the public is not adversely affected. The proposed relocation
retains the current TS requirements and only alters the location of
these provisions. This relocation cannot affect the probability or
consequences of an accident as this is only an administrative
revision that will not alter any plant equipment or processes.
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.
Since the proposed change only relocates the current TS
requirements without change, there is not a potential for a change
in the accident generation potential. This change will not alter
plant components, systems, or operating practices. 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 relocates specifications that do not meet
the threshold for inclusion in the TSs as defined in 10 CFR 50.36.
This change will not alter the requirements for these functions or
plant setpoints or functions that maintain the margins 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consderation Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the
[[Page 38724]]
action involved exigent circumstances. They are repeated here because
the biweekly notice lists all amendments issued or proposed to be
issued involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Brunswick County, North Carolina
Date of amendment request: May 17, 2005.
Brief description of amendment request: The amendments replace the
existing requirement of Technical Specification 3.4.5, ``RCS [Reactor
Coolant System] Leakage Detection Instrumentation,'' Required Action
D.1, to enter Limiting Condition for Operation (LCO) 3.0.3 if required
leakage detection systems are inoperable with the requirement to be in
Mode 3 within 12 hours and Mode 4 within 36 hours.
Date of publication of individual notice in Federal Register: June
13, 2005 (70 FR 34161).
Expiration date of individual notice: June 27, 2005 (for comments);
August 12, 2005 (for hearing requests).
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (PDR), located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by email to [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: February 24, 2005.
Brief description of amendment: The amendment revised the Technical
Specifications, Section 3.1.1, ``Protective Instrumentation
Requirements,'' notes aa and bb, correcting missed wording which led to
incorrect statements of the as-designed service water pump and reactor
building closed cooling water system pump trip conditions. The
amendment also made an editorial correction to pages 3.6-1 and 3.6-2.
Date of Issuance: June 23, 2005.
Effective date: June 23, 2005 and shall be implemented within 60
days of issuance.
Amendment No.: 255.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 29, 2005 (70 FR
15941). The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated June 23, 2005.
No significant hazards consideration comments received: No.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of application for amendment: October 21, 2004, as
supplemented January 4, 2005.
Brief description of amendment: The amendment deletes the Technical
Specification (TS) requirements to submit monthly operating reports and
annual occupational radiation exposure reports. The change is
consistent with Revision 1 of the Nuclear Regulatory Commission (NRC)
approved Industry/Technical Specifications Task Force (TSTF) Standard
Technical Specification Change Traveler, TSTF-369, ``Removal of Monthly
Operating Report and Occupational Radiation Exposure Report.'' This TS
improvement was published in the Federal Register on June 23, 2004 (69
FR 35067), as part of the Consolidated Line Item Improvement Process.
Date of issuance: June 17, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 254.
Facility Operating License No. DPR-50. Amendment revised the TSs.
Date of initial notice in Federal Register: April 12, 2005 (70 FR
19114). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 17, 2005.
No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: July 13, 2004, as supplemented
on April 21, 2005.
Brief description of amendments: The amendments revised License
Condition 2.E of each unit's operating license by replacing the current
wording with wording from Generic Letter (GL) 86-10, ``Implementation
of Fire Protection Requirements.''
Date of issuance: June 15, 2005.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 273 and 250.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the operating licenses.
Date of initial notice in Federal Register: December 7, 2004 (69 FR
70715). The supplement dated April 21, 2005, provided additional
information that clarified the application, did not expand the scope of
the application as
[[Page 38725]]
originally noticed, and did not change the staff's original proposed no
significant hazards consideration determination. The Commission's
related evaluation of these amendments is contained in a Safety
Evaluation dated June 15, 2005.
No significant hazards consideration comments received: No.
Carolina Power & Light Company, Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: January 27, 2005.
Brief Description of amendments: The amendments revised respective
Technical Specifications (TS) testing frequency for the surveillance
requirement (SR) in TS 3.1.4, ``Control Rod Scram Times.'' The change
revises the test frequency of SR 3.1.4.2, control rod scram time
testing, from ``120 days cumulative operation in MODE 1'' to ``200 days
cumulative operation in MODE 1.''
Date of issuance: May 31, 2005.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment Nos.: 236 and 264.
Facility Operating License Nos. DPR-71 and DPR-62: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: February 15, 2005 (70
FR 12745). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 31, 2005.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., Docket Nos. 50-245, 50-336, and 50-
423, Millstone Power Station, Unit Nos. 1, 2, and 3, New London County,
Connecticut
Date of application for amendments: December 21, 2004.
Brief description of amendments: The amendments eliminate
requirements for annual Occupational Radiation Exposure Reports, annual
reports regarding challenges to pressurizer relief and safety valves,
and Monthly Operating Reports.
Date of issuance: June 13, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 114, 286, and 223.
Facility Operating License Nos. DPR-21, DPR-65, and NPF-49: The
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 12, 2005 (70 FR
19114). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated June 13, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: March 22, 2004, as supplemented
by letters dated February 8 and April 7, 2005.
Brief description of amendments: The amendments revised the
Technical Specifications (TS) 3.3.2, ``Engineered Safety Features
Actuation System Instrumentation,'' and TS 3.3.6, ``Containment Air
Release and Addition Isolation Instrumentation,'' to permit an 18-month
surveillance interval for certain Westinghouse Type AR slave relays and
for certain Potter and Brumfield MDR-Series slave relays.
Date of issuance: May 24, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 224 and 219.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 14, 2004 (69
FR 55468). The supplements dated February 8 and April 7, 2005, provided
additional information that clarified the application, did not expand
the scope of the March 22, 2004, application as originally noticed, and
did not change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 24, 2005.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: June 10, 2004, as supplemented
by letter dated January 31, 2005.
Brief description of amendments: The amendments revised the
Technical Specifications to extend the interval between local leakage
rate tests of the containment purge and vent valves with resilient
seals (that is, in the containment purge system, hydrogen purge system,
and containment air release and addition system).
Date of issuance: June 10, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 225 and 222.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 21, 2004 (69
FR 76487).
The supplement dated January 31, 2005, provided additional
information that clarified the application, did not expand the scope of
the June 10, 2004, application as originally noticed, and did not
change the staff's original proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 10, 2005.
No significant hazards consideration comments received: No
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
2, Pope County, Arkansas
Date of amendment request: May 12, 2004, as completely superseded
by application dated July 8, 2004, and supplemented by letters dated
October 14, 2004, and January 19, March 7, and April 7, 2005.
Brief description of amendment: The Index is deleted from the
Technical Specifications.
Date of issuance: June 22, 2005.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 260.
Facility Operating License No. NPF-6: Amendment deletes the
Technical Specifications Index.
Date of initial notice in Federal Register: August 31, 2004 (69 FR
53106). The supplements dated October 14, 2004, and January 19, March
7, and April 7, 2005, 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 as published in the
Federal Register.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 22, 2005.
No significant hazards consideration comments received: No.
[[Page 38726]]
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: October 21, 2004.
Brief description of amendments: The amendment deletes the
Technical Specification (TS) requirements to submit monthly operating
reports and annual occupational radiation exposure reports. The change
is consistent with Revision 1 of NRC-approved Technical Specifications
Task Force (TSTF) 369, ``Elimination of Requirements for Monthly
Operating Reports and Occupational Radiation Exposure Reports.'' This
TS improvement was published in the Federal Register (69 FR 35067) on
June 23, 2004, as part of the Consolidated Line Item Improvement
Process.
Date of issuance: June 14, 2005.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendments Nos.: 254 and 257.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 12, 2004 (70 FR
19116). The Commission's related evaluation of the amendments are
contained in a Safety Evaluation dated June 14, 2005.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad
Cities Nuclear Power Station, Units 1 and 2, Rock Island County,
Illinois
Date of application for amendments: September 15, 2004.
Brief description of amendments: The amendments deleted the
technical specification (TS) requirements related to hydrogen and
oxygen monitors. The TS changes support implementation of the revisions
to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.44,
``Combustible Gas Control for Nuclear Power Reactors,'' that became
effective on October 16, 2003. The changes are consistent with Revision
1 of the NRC-approved Industry/Technical Specifications Task Force
(TSTF) Standard Technical Specification Change Traveler, TSTF-447,
``Elimination of Hydrogen Recombiners and Change to Hydrogen and Oxygen
Monitors.''
Date of issuance: June 14, 2005.
Effective date: June 14, 2005.
Amendment Nos.: 226/221.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 2005 (70 FR
5243).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 14, 2005.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: August 2, 2004.
Brief description of amendment: This amendment deleted Technical
Specification 6.8.4.c, ``Post-Accident Sampling,'' and the related
requirements to maintain a Post-Accident Sampling System.
Date of issuance: June 10, 2005.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 264.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 12, 2004 (69 FR
60682).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 10, 2005.
No significant hazards consideration comments received: No.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: October 22, 2004, as supplemented by
letter dated December 16, 2004.
Description of amendment request: The amendment revised the
Seabrook Station, Unit No. 1 Technical Specifications (TSs) to allow
for individual entry into the limiting condition for operation (LCO)
for each instrument, and extends the allowed outage times for LCOs
3.3.3.6.a and 3.3.3.6.b.
Date of issuance: June 15, 2005.
Effective date: As of its date of issuance, and shall be
implemented within 30 days.
Amendment No.: 103.
Facility Operating License No. NPF-86: The amendment revised the
TSs.
Date of initial notice in Federal Register: November 2, 2004 (69 FR
63560). The December 16, 2004 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 June 15, 2005.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: May 5, 2005, as supplemented
June 9, 2005.
Brief description of amendment: The amendment revises the Facility
Operating License and Technical Specifications to modify the auxiliary
feed water (AFW) pump suction protection requirements and change the
design basis as described in the Updated Safety Analysis Report to
revise the functionality of the discharge pressure switches to provide
pump runout protection, which requires operator actions to restore the
AFW pumps for specific post-accident recovery activities.
Date of issuance: June 20, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 183.
Facility Operating License No. DPR-43: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: May 13, 2005 (70 FR
25619). The supplement dated June 9, 2005, provided clarifying
information that did not change the scope of the May 5, 2005
application, nor the initial proposed no significant hazards
consideration determination as published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 20, 2005.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: September 30, 2004.
Brief description of amendments: The amendments revise Technical
Specifications related to the reactor coolant pump flywheel inspection
program by increasing the inspection interval from current 10 years to
20 years.
Date of issuance: June 10, 2005.
[[Page 38727]]
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 118/118.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 1, 2005 (70 FR
9998).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 10, 2005.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of application for amendment: September 12, 2003, as
supplemented by letters dated November 20, 2003, March 30, April 20,
May 7, May 27, August 18, and November 3, 2004, and February 17, 2005.
Brief description of amendment: These amendments revise the
Technical Specifications to incorporate a full-scope application of an
alternate source term methodology in accordance with Title 10 of the
Code of Federal Regulations, Section 50.67.
Date of issuance: June 15, 2005.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 240 and 221.
Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: December 9, 2003 (68 FR
68672). The supplements dated November 20, 2003, March 30, April 20,
May 7, May 27, August 18, and November 3, 2004, and February 17, 2005,
contained clarifying information only and did not change the initial no
significant hazards consideration determination or expand the scope of
the initial application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 15, 2005.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units 1 and 2, Louisa County, Virginia
Date of application for amendment: August 30, 2004.
Brief description of amendment: These amendments revise the
Technical Specifications by extending the inspection interval for
reactor coolant pump flywheels to 20 years.
Date of issuance: June 15, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 241 and 222.
Renewed Facility Operating License Nos. NPF-4 and NPF-7: Amendments
change the Technical Specifications.
Date of initial notice in Federal Register: March 15, 2005 (70 FR
12751).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 15, 2005.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: August 30, 2004.
Brief Description of amendments: These amendments revise the
Technical Specifications by extending the inspection interval for
reactor coolant pump flywheels to 20 years.
Date of issuance: June 21, 2005.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 242 and 241.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 15, 2005 (70 FR
12751).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 21, 2005.
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 Announcment 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.
[[Page 38728]]
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 email to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. Within 60 days after the date
of publication of this notice, the licensee may file a request for a
hearing with respect to issuance of the amendment to the subject
facility operating license and any person whose interest may be
affected by this proceeding and who wishes to participate as a party in
the proceeding must file a written request for a hearing and a petition
for leave to intervene. Requests for a hearing and a petition for leave
to intervene shall be filed in accordance with the Commission's ``Rules
of Practice for Domestic Licensing Proceedings'' in 10 CFR Part 2.
Interested persons should consult a current copy of 10 CFR 2.309, which
is available at the Commission's PDR, located at One White Flint North,
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville,
Maryland, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or a presiding officer designated by the
Commission or by the Chief Administrative Judge of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the Chief Administrative Judge of the Atomic
Safety and Licensing Board will issue a notice of a hearing or an
appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also identify the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or
fact.\1\ Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner/requestor
who fails to satisfy these requirements with respect to at least one
contention will not be permitted to participate as a party.
---------------------------------------------------------------------------
\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and
Adjudications Staff; (3) E-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile
[[Page 38729]]
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 email 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).
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: June 16, 2005, as supplemented June 19,
2005.
Description of amendment request: The amendment revises the
Technical Specifications to remove the requirement to have an operable
containment spray flow path capable of taking suction from the
containment sump.
Date of issuance: June 21, 2005.
Effective date: June 21, 2005.
Amendment No.: 184.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No. The Commission's related evaluation of the
amendment, finding of emergency circumstances, state consultation, and
final NSHC determination are contained in a safety evaluation dated
June 21, 2005.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P.O. Box 1497, Madison, WI 53701-1497.
NRC Section Chief: L. Raghavan.
Dated in Rockville, Maryland, this 27th day of June 2005.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 05-12987 Filed 7-1-05; 8:45 am]
BILLING CODE 7590-01-P