[Federal Register Volume 69, Number 138 (Tuesday, July 20, 2004)]
[Notices]
[Pages 43457-43465]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-16157]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
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 to 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 25, 2004, through July 8, 2004. The
last biweekly notice was published on July 6, 2004 (69 FRN 40668).
[[Page 43458]]
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address and telephone
number of the requestor or petitioner; 2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of
[[Page 43459]]
the amendment. If the final determination is that the amendment request
involves a significant hazards consideration, any hearing held would
take place before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express mail, and expedited delivery services:
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852, Attention: Rulemaking and
Adjudications Staff; (3) e-mail addressed to the Office of the
Secretary, U.S. Nuclear Regulatory Commission, [email protected];
or (4) facsimile transmission addressed to the Office of the Secretary,
U.S. Nuclear Regulatory Commission, Washington, DC, Attention:
Rulemakings and Adjudications Staff at (301) 415-1101, verification
number is (301) 415-1966. A copy of the request for hearing and
petition for leave to intervene should also be sent to the Office of
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and it is requested that copies be transmitted either by
means of facsimile transmission to (301) 415-3725 or by e-mail to
[email protected]. A copy of the request for hearing and petition
for leave to intervene should also be sent to the attorney for the
licensee.
Nontimely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the 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
NRC PDR Reference staff at 1-800-397-4209, (301) 415-4737 or by e-mail
to [email protected].
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: May 25, 2004.
Description of amendment request: The proposed amendments would
revise the licensing basis in the Updated Final Safety Analysis Report
to support installation of a passive low-pressure injection (LPI) cross
connect inside containment for Unit 3. The proposed changes would
revise the licensing basis for selected portions of the core flood and
LPI piping to allow exclusion of the dynamic effects associated with a
postulated rupture of that piping by application of leak-before-break
technology. Similar amendments were approved for Unit 1 by NRC letter
dated September 29, 2003, and for Unit 2 by NRC letter dated February
5, 2004.
The proposed amendments would also delete technical specifications
(TSs) which will no longer apply when the LPI cross connect
modification has been implemented.
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) Involve a significant increase in the probability or
consequences of an accident previously evaluated: The proposed
License Amendment Request (LAR) modifies the Unit 3 licensing basis
to allow the dynamic effects associated with postulated pipe rupture
of selected portions of the Unit 3 Low Pressure Injection (LPI)/Core
Flood (CF) piping to be excluded from the design basis. The proposed
LAR also removes Technical Specifications that are no longer
applicable due to the completion of the LPI cross connect
modification on all three Oconee Units. The proposed design
allowances for these selected portions of piping continue to allow
the LPI system design to meet General Design Criteria (GDC) 4
requirements related to environmental and dynamic effects. The
proposed LAR will continue to ensure that ONS [Oconee Nuclear
Station] can meet design basis requirements associated with the LPI
safety function. The addition of the crossover line will enhance the
ability of the control room operator to mitigate the consequences of
specific events for which LPI is credited. Therefore, the proposed
LAR does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated: The
proposed LAR modifies the Unit 3 licensing basis to allow the
dynamic effects associated with postulated pipe rupture of selected
portions of Unit 3 LPI/CF piping to be excluded from the design
basis and removes TS requirements that are no longer applicable due
to the completion of the LPI cross connect modification on all three
Oconee Units. The proposed design allowances for these selected
portions of piping continue to allow the LPI system design to meet
GDC 4 requirements related to environmental and dynamic effects. The
systems affected by the changes are used to mitigate the
consequences of an accident that has already occurred. The proposed
licensing basis change does not affect the mitigating function of
these systems. Consequently, these changes do not alter the nature
of events postulated in the Safety Analysis Report nor do they
introduce any unique precursor mechanisms. Therefore, the proposed
amendment will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
(3) Involve a significant reduction in a margin of safety: The
proposed licensing basis and TS changes do not unfavorably affect
any plant safety limits, set points, or design parameters. The
changes also do not unfavorably affect the fuel, fuel cladding, RCS
[Reactor Coolant System], or containment integrity. Therefore, the
proposed changes, which add new design allowances associated with
the passive LPI cross connect modification and remove obsolete TS
requirements, 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: Anne W. Cottingham, Winston and Strawn LPP,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Stephanie M. Coffin (Acting).
Entergy Nuclear Operations, Inc., Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant (JAFNPP), Oswego County, New York
Date of amendment request: June 4, 2004.
Description of amendment request: The proposed amendment would
revise the safety limit values in Technical Specification (TS) 2.1.1.2
for the minimum critical power ratio (MCPR) for both single and two
recirculation loop operation.
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:
[[Page 43460]]
1. The operation of JAFNPP in accordance with the proposed
amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The basis of the Safety Limit Minimum Critical Power Ratio
(SLMCPR) is to ensure no mechanistic fuel damage is calculated to
occur if the limit is not violated. The new SLMCPR values preserve
the existing margin to transition boiling and probability of fuel
damage is not increased. The derivation of the revised SLMCPR for
JAFNPP for incorporation into the Technical Specifications, and its
use to determine plant and cycle-specific thermal limits, have been
performed using NRC approved methods. These plant-specific
calculations are performed each operating cycle and if necessary,
will require future changes to these values based upon revised core
designs. The revised SLMCPR values do not change the method of
operating the plant and have no effect on the probability of an
accident initiating event or transient.
Based on the above, JAFNPP has concluded that the proposed
change will not result in a significant increase in the probability
or consequences of an accident previously evaluated.
2. The operation of JAFNPP in accordance with the proposed
amendment, will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed changes result only from a specific analysis for
the JAFNPP core reload design. These changes do not involve any new
or different methods for operating the facility. No new initiating
events or transients result from these changes.
Based on the above, JAFNPP has concluded that the proposed
change will not create the possibility of a new or different kind of
accident from those previously evaluated.
3. The operation of JAFNPP in accordance with the proposed
amendment, will not involve a significant reduction in a margin of
safety.
The new SLMCPR is calculated using NRC approved methods with
plant and cycle specific parameters for the current core design. The
SLMCPR value remains high enough to ensure that greater than 99.9%
of all fuel rods in the core will avoid transition boiling if the
limit is not violated, thereby preserving the fuel cladding
integrity. The operating MCPR limit is set appropriately above the
safety limit value to ensure adequate margin when the cycle specific
transients are evaluated. Accordingly, the margin of safety is
maintained with the revised values.
As a result, JAFNPP has determined that the proposed change will
not result in 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. David E. Blabey, 1633 Broadway, New
York, New York 10019.
NRC Section Chief: Richard J. Laufer.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: June 17, 2004.
Description of amendment request: The amendment will (1) modify
Technical Specifications (TSs) 5.3.1, Fuel Assemblies, to allow a
limited number of lead test assemblies (LTAs) and limited substitutions
of zirconium alloy or stainless steel filler rods for fuel rods, (2)
include ZIRLOTM as an acceptable fuel rod cladding which is
consistent with 10 CFR 50.46, (3) relocate some of the information in
TS 5.3.1 to TS 5.6.1, (4) change TS 6.9.1.11.1 to allow the use of the
Westinghouse Nuclear Physics code package and to incorporate the
methodology used to support ZIRLOTM cladding material, and
(5) delete the Index from the TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
TS 5.3.1, Fuel Assemblies and TS 5.6.1, Criticality
The proposed change allows the use of a limited number of lead
test assemblies; the use of limited substitutions of zirconium alloy
or stainless steel filler rods for fuel rods; and the use of methods
required for the implementation of ZIRLOTM clad fuel
rods. Inasmuch as the revision identifies codes previously approved
by the NRC [Nuclear Regulatory Commission] for CE [Combustion
Engineering] cores, the amendment is administrative in nature and
has no impact on any plant configuration or system performance
relied upon to mitigate the consequences of an accident.
The proposed change in part represents a relocation of a portion
of the information previously located in the TSs design features
section to the FSAR [Final Safety Analysis Report], which is
controlled under 10 CFR 50.59, ``Changes, Tests, and Experiments.''
This change is administrative in nature because the design
requirements for the facility remain the same.
The proposed change does not remove or modify any of the design
requirements for the facility or affect any accident initiators,
conditions or assumption[s] for an accident previously evaluated.
TS 6.9.1.11, Core Operating Limits Report COLR
The proposed amendment identifies a change in the nuclear
physics codes used to confirm the values of selected cycle-specific
reactor physics parameter limits and includes minor editorial
changes which do not alter the intent of stated requirements. The
proposed change also allows the use of methods required for the
implementation of ZIRLOTM clad fuel rods. Inasmuch as the
proposed change identifies codes previously approved by the NRC for
CE cores, the amendment is administrative in nature and has no
impact on any plant configuration or system performance relied upon
to mitigate the consequences of an accident. Parameter limits
specified in the site specific COLR are not changed from the values
presently required by TSs. Future changes to the calculated values
of such limits may only be made using NRC approved methodologies,
must be consistent with all applicable safety analysis limits, and
are controlled by the 10 CFR 50.59 process. Assumptions used for
accident initiators and/or safety analysis acceptance criteria are
not changed by this change.
Index
The proposed change is administrative in nature and does not
affect any system or component functional requirements. This change
does not affect the operation of the plant or affect any component
that is used to mitigate the consequences of any accident.
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.
TS 5.3.1, Fuel Assemblies and TS 5.6.1, Criticality
The proposed change allows the use of methods required for the
implementation of ZIRLOTM clad fuel rods. Inasmuch as the
revision identifies codes previously approved by the NRC for CE
cores, the amendment is administrative in nature and has no impact
on any plant configuration or system performance relied upon to
mitigate the consequences of an accident.
In addition, the proposed change allows the use of a limited
number of lead test assemblies. The proposed change is
administrative in nature. Prior to the use of lead test assemblies,
fuel designs will be analyzed with applicable NRC staff approved
codes and methods and shown by tests or analyses to comply with all
fuel safety design bases to assure no new or different kind of
accident from any accident previously evaluated will be created.
And finally the proposed change allows the relocation of a
portion of the information previously located in the TSs design
features section to the FSAR. This change is administrative in
nature and does not create a new or different type of accident than
previously evaluated because the design requirements for the
facility remain the same.
[[Page 43461]]
The proposed change does not remove or modify any of the design
requirements for the facility or affect any accident initiators,
conditions or assumption[s] for an accident previously evaluated.
TS 6.9.1.11, Core Operating Limits Report COLR
The proposed change identifies a change in the Nuclear Physics
codes used to confirm the values of selected cycle-specific reactor
physics parameter limits contained in the COLR. The proposed change
also allows the use of methodologies required for the implementation
of ZIRLOTM clad fuel rods. Neither of these changes
results in a change [to] the physical plant or the modes of
operation defined in the facility license.
Index
The proposed change is administrative in nature and does not
affect any system or component functional requirements. This change
does not affect the operation of the plant or affect any component
that is used to mitigate the consequences of any accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
TS 5.3.1, Fuel Assemblies and TS 5.6.1, Criticality
The proposed change allows the use of methods required for the
implementation of ZIRLOTM clad fuel rods. Inasmuch as the
revision identifies codes previously approved by the NRC for CE
cores, the amendment is administrative in nature and has no impact
on any plant configuration or system performance relied upon to
mitigate the consequences of an accident.
In addition, the proposed change allows the use of a limited
number of lead test assemblies. The proposed change is
administrative in nature. Prior to the use of lead test assemblies,
fuel designs will be analyzed with applicable NRC staff approved
codes and methods and shown by tests or analyses to ensure
compliance with any safety analysis acceptance criteria.
And finally the proposed change allows the relocation of a
portion of the information previously located in the TSs design
features section to the FSAR. This change is administrative in
nature and does not create a new or different type of accident than
previously evaluated because the design requirements for the
facility remain the same.
The proposed change does not remove or modify any of the design
requirements for the facility or affect any accident initiators,
conditions or assumption[s] for an accident previously evaluated.
TS 6.9.1.11, Core Operating Limits Report COLR
The individual specifications continue to require operation of
the plant within the bounds of the limits specified in COLR.
Benchmarking has shown that uncertainties for the Westinghouse
Physics code system (ANC/PHOENIX-P) yields are essentially the same
or less than those obtained for the current ROCS/DIT methodology.
Future changes to the values of these limits by the licensee may
only be developed using NRC approved methodologies, remaining
consistent with all applicable plant safety analysis limits
addressed in the Safety Analysis Report, which are controlled by the
10 CFR 50.59 process. The relocation of the supplement numbers,
revision numbers, and approval dates related to the analytical
methods listed in the COLR does not affect the margin of safety. The
analysis will continue to be performed using NRC approved
methodology. Safety analysis acceptance criteria are not being
altered by this change.
Index
The proposed change is administrative in nature and does not
affect any system or component functional requirements. Safety
analysis acceptance criteria are not being altered by this change.
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: N.S. Reynolds, Esquire, Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Dockets 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: June 15, 2004.
Description of amendment request: The proposed amendment would
allow the licensee to conduct the monthly diesel surveillance test, the
diesel full-load rejection test, the diesel 24-hour run test and the
diesel hot restart test at the higher load of 2800 kW.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed revisions to Technical Specification [TS]
Surveillance Requirements SR 3.8.1.3 (the monthly diesel
surveillance test), SR 3.8.1.10 (the diesel full-load rejection
test), SR 3.8.1.14.b (the diesel 24-hour run test), and SR 3.8.1.15
(the diesel hot restart test) to permit these tests to be conducted
at the higher load value of 2800 kW do not involve any physical
change to any EDG [emergency diesel generator] equipment. The
Operator using existing EDG load controls will adjust the EDG to
carry the increased load during surveillance testing.
The EDGs are designed to provide a reliable source of AC
electrical power in the event of an accident coincident with a loss
of offsite power. The failure of an EDG itself is not considered an
accident evaluated in the UFSAR [Updated Final Safety Analysis
Report]. This proposed loading change does not affect the current
accident initiators or precursors that could lead to a previously
evaluated accident.
The failure of a single EDG to perform when required to mitigate
the consequences of an accident has already been considered as a
subsequent single failure in the current plant safety analyses. The
proposed change to increase the allowable load range does not alter
the EDG design features, post-accident operation, or accident
analysis assumptions which could affect the ability of the EDGs to
mitigate the consequences of a previously evaluated accident.
Current EDG testing requirements, e.g., starting, timing, and post
accident sequencing and loading will continue to ensure reliable EDG
operation and are not being changed in this request.
Since the EDG TS surveillance test load is the only parameter
involved in this request, the proposed changes will not increase the
likelihood of the malfunction of another system, structure, or
component that has been assumed as an accident initiator or credited
in the mitigation of an accident.
Based on the above discussion, the proposed TS changes do 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 EDGs are designed to provide a reliable source of AC
electrical power in the event of an accident coincident with a loss
of offsite power. No change in the ability of the EDGs to perform
their design function is involved. Instrumentation setpoints,
starting, sequencing, and post-accident loading functions associated
with the EDGs are not affected by the proposed changes. No
modifications to the EDGs are required to implement the proposed TS
changes. Therefore, no new failure mechanism, malfunction, or
accident initiator is considered credible.
Additionally, the proposed TS changes do not affect the other
plant design, hardware, system operation, or procedures. Therefore,
based on the above discussion, the above TS changes do 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 underlying purpose of the four (4) diesel generators is to
ensure an available source of onsite power to the ESF [engineered
safety feature] systems. This
[[Page 43462]]
change does [sic] will not impact this underlying purpose. As
discussed above, this change may result in a slight increase in
engine wear due to the ability to operate at the higher load, but
this increased wear is bounded by the existing 24 month maintenance
inspection program. The OEM [original equipment manufacturer] has
stated that the change to increase the allowable load value still
remains well within the EDG 2000-hour rating, and the increased rate
of wear is within the acceptable limits of the current maintenance
program.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has found that, because the EDGs will continue to be
operated within the bounds of the current maintenance program, there is
no significant increase in the probability of an EDG failure;
therefore, there is no significant increase in the probability or
consequences of an accident previously evaluated. The NRC staff further
finds that, because there is no significant increase in a failure of an
EDG to perform its function, the proposed change does not create the
possibility of an accident not previously evaluated.
The NRC staff has reviewed the licensee's analysis and, based on
this review and the staff's own findings above, it appears that the
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC
staff proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for Licensee: Thomas S. O'Neill, Associate and General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Section Chief: James W. Clifford.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and 2),
Beaver County, Pennsylvania
Date of amendment request: June 2, 2004.
Description of amendment request: The proposed amendment would
revise the BVPS-1 and 2 Technical Specifications to allow operation
with atmospheric containment designs.
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 Beaver Valley Power Station (BVPS)
containments are designed to withstand the internal pressure and
temperature resulting from a loss of coolant accident (LOCA), main
steamline break (MSLB), feedwater line break, and a control rod
ejection accident (CREA). Each of these accidents has been
previously analyzed with the results provided in the Updated Final
Safety Analysis Report (UFSAR) except the feedwater line break. This
accident is not analyzed because the MSLB is more limiting. The
affect on containment pressure and temperature due to a CREA is
bounded by a LOCA, since a CREA is modeled after a small break LOCA.
The probability of occurrence for these accidents is independent of
the type of containment. Additionally the supporting plant
modifications will not increase the probability of an accident
because they perform an accident mitigation function and are not
accident initiators. Therefore a change from sub-atmospheric to an
atmospheric containment will not increase the probability of these
accidents.
For accident conditions, the proposed changes will potentially
impact the reported dose consequences of the LOCA and CREA for both
BVPS units. The radiological consequences of these and the remaining
design basis accidents are not adversely impacted by the proposed
changes because they are within the current BVPS licensing and
design basis.
From a containment integrity viewpoint, the limiting DBA
[design-basis accident] presently is the MSLB for Unit 1 and the
LOCA for Unit 2. Following the conversion to an atmospheric
containment the limiting DBA will be the LOCA for both units. The
revised containment integrity analysis demonstrates that with the
installation of the supporting plant modifications that the
pressures and temperatures associated with the applicable design
basis accidents identified above are within the existing containment
design limits.
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 design basis accidents, which could be
adversely affected by the proposed changes, have been reanalyzed.
These [re]analyses demonstrate that all acceptance criteria have
been satisfied. The revised containment integrity analysis
demonstrates that the containment will not be subjected to
temperatures or pressures that are beyond its design limits.
Converting to an atmospheric containment will not result in any new
or different kind of accidents because no new accident initiators
will be introduced.
The affects of the supporting plant modifications and the
proposed Technical Specification changes on plant structures,
systems and components (SSC) have been evaluated and it has been
verified that the capability of the SSCs to perform their design
functions will be retained following approval of the proposed
Technical Specification changes and installation of the supporting
plant modifications.
Changes to instrumentation setpoints, surveillance requirements,
installation of the supporting plant modifications, and the
elimination of certain operability requirements will not create the
possibility of a new or different type of accident since these
changes would not result in significant changes to the manner in
which the affected equipment is operated during normal plant
operations.
Therefore, 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 margin of safety attributed to the containment
involves both the pressures and temperatures the containment is
subjected to following a DBA, and the on-site and offsite dose
consequences associated with normal and post DBA operations.
The revised containment analyses demonstrates that, following a
DBA; containment peak pressure and temperature will not exceed the
containment's design limits and that the containment pressure will
not decrease to below 8 psia following the intentional or
inadvertent actuation of the quench spray system. Since the
containment design limits are not exceeded, the existing margin of
safety between these limits and the containment failure limits is
not reduced.
Since the current radiological analyses impacted by the
containment conversion are conservatively based on atmospheric
operation, it is concluded that the existing dose consequence margin
of safety will not be impacted when the BVPS units are operated with
an atmospheric containment.
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 O'Reilly, FirstEnergy Nuclear Operating
Company, FirstEnergy Corporation, 76 South Main Street, Akron, OH
44308.
NRC Section Chief: Richard J. Laufer.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of amendment request: April 26, 2004.
Description of amendment request: The proposed amendments would
revise Technical Specifications Limiting Conditions for Operation (LCO)
3.7.9, ``Ultimate Heat Sink (UHS)'' to allow the UHS to remain OPERABLE
with three of four fans operating under certain environmental
conditions.
[[Page 43463]]
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 consequence of an accident previously evaluated?
No. The revised requirements will maintain OPERABILITY while
allowing maintenance on one fan when ambient wet-bulb temperature is
63 [deg]F or lower. Modifying the condition when one NSCW [nuclear
service cooling water] tower is impacted is more restrictive. The
UHS is not an initiator to any analyzed accident sequence. Operation
in accordance with the proposed TS will continue to ensure that the
UHS remains capable of performing its safety function and that all
analyzed accidents will continue to be mitigated as previously
analyzed. 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 previously evaluated?
No. The proposed changes do not introduce any new equipment,
create new failure modes for existing equipment, or create any new
limiting single failures. Plant operation will not be altered, and
all safety functions previously addressed in accident analyses will
continue to be performed. 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?
No. The proposed changes will not adversely affect operation of
plant equipment-principally the UHS and the equipment supported by
it. Modifying the condition where one NSCW tower is impacted is more
restrictive and enhances the margin of safety. Therefore, the
proposed changes do not involve a significant reduction in any
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. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: Stephanie M. Coffin (Acting).
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 21, 2004.
Description of amendment request: The proposed one-time (per unit)
change revises the steam generator (SG) inservice inspection frequency
requirements in Technical Specification (TS) 4.4.5.3a for Unit 1
immediately after the tenth refueling outage for Unit 1 (1RE10) and for
Unit 2 immediately after refueling outage 2RE10. The change would allow
a 78-month inspection interval after one inspection resulting in C-1
classification, rather than a 40-month interval after two consecutive
inspections resulting in C-1 classification.
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.
There is no direct increase in SG leakage because the proposed
change does not alter the plant design. The scope of inspections
performed during 1RE10 and 2RE10, the first refueling outage
following SG replacement, exceeded the combined TS requirements for
the first two refueling outages after replacement. That is, more
tubes were inspected than were required by TS. Currently, neither
Unit 1 nor Unit 2 has an active SG damage mechanism and will meet
the current industry examination guidelines without performing
inspections during the next 78 months. The Condition Monitoring
Assessment after 1RE10 and 2RE10 demonstrated that all performance
criteria were met during these outages. The Operational Assessment
shows that all performance criteria will be met over the proposed
operating period.
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 change will not alter any plant design basis or
postulated accident resulting from potential SG tube degradation.
The scope of inspections performed during 1RE10 and 2RE10, the first
refueling outage for each unit following SG replacement,
significantly exceeded the combined TS requirements for the scope of
the first two refueling outages after SG replacement. The
inspections already performed exceed those required by the current
TS over the proposed 78-month period.
The proposed change does not affect the design of the SGs, the
method of operation, or reactor coolant chemistry controls. No new
equipment is being introduced and installed and equipment is not
being operated in a new or different manner. The proposed change
involves a one-time extension of the SG tube inservice inspection
interval, and therefore will not give rise to new failure modes. In
addition, the proposed change does not impact any other plant system
or components.
Therefore, 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.
Steam generator tube integrity is a function of design,
environment, and current physical condition. Extending the SG tube
inservice inspection interval to 78 months will not alter the
function or design of the SGs. Inspections conducted prior to
placing the SGs into service (pre-service inspections) and
inspection during the first refueling outages following SG
replacement demonstrate that the SGs do not have fabrication damage
or an active damage mechanism. The scope of those inspections
significantly exceeded those required by the TS. These inspection
results were comparable to similar inspection results for the same
model of RSGs [replacement steam generators] installed at other
plants, and subsequent inspections at those plants yielded results
that support this extension request. The improved design of the RSGs
also provides reasonable assurance that significant tube degradation
is not likely to occur over the proposed operating period.
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 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: Robert A. Gramm.
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
[[Page 43464]]
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, 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 NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, (301) 415-4737 or by e-mail to [email protected].
Dominion Nuclear Connecticut, Inc., Docket No. 50-245, Millstone Power
Station, Unit No. 1, New London County, Connecticut
Date of amendment request: September 18, 2003.
Brief description of amendment: The amendment revises Technical
Specification 4.2, ``Fuel Storage,'' to eliminate all credit for
Boraflex as a neutron absorber, reduce the number of fuel assemblies
allowed to be stored in the spent fuel pool (SFP), change the required
SFPkeff and eliminate design features requirements of new
fuel storage.
Date of issuance: June 29, 2004.
Effective date: June 29, 2004, and shall be implemented within 60
days from the date of issuance.
Amendment No.: 113.
Facility Operating License No. DPR-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 9, 2003 (68 FR
68659). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 29, 2004.
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: October 15, 2003.
Brief description of amendments: The amendments added a new
Technical Specification (TS) 3.9.7, ``Unborated Water Source isolation
Valves,'' and revised TS 3.9.2, ``Nuclear Instrumentation,'' to delete
the requirement for Boron Dilution Mitigation System automatic valve
actuations and makeup water pump trip during Mode 6 and to agree with
the wording of NUREG-1431, ``Standard Technical Specifications
Westinghouse Plants,'' Revision 2. The licensee proposed these changes
to provide configuration control of the dilution valves during Mode 6
to preclude the possibility of a boron dilution event and to provide an
opportunity to conduct maintenance on the volume control tank valves,
refueling water storage tank valves, and their respective power
supplies.
Date of issuance: June 21, 2004.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 215 and 209.
Renewed Facility Operating License Nos. NPF-35 and NPF-52:
Amendments revised the TSs.
Date of initial notice in Federal Register: March 16, 2004 (69 FR
12366).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 21, 2004.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: June 25, 2003.
Brief description of amendments: The amendments are administrative
in nature and incorporate several editorial changes.
Date of issuance: June 21, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 222 and 204.
Renewed Facility Operating License Nos. NPF-9 and NPF-17:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19565).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 21, 2004.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Docket Nos. 50-247 and 50-286, Indian Point
Nuclear Generating Unit Nos. 2 and 3, Westchester County, New York
Date of application for amendment: March 3, 2004.
Brief description of amendments: The amendments revised the
Technical Specifications administrative controls requirements regarding
the reactor coolant pump flywheel inspection program to increase the
inspection interval from 10 years to 20 years.
Date of issuance: July 2, 2004.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 240 and 221.
Facility Operating License Nos. DPR-26 and DPR-64: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19566).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 2, 2004.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: August 27, 2003, as
supplemented December 15, 2003, and February 27, 2004.
Brief description of amendments: The amendments modify Technical
Specifications requirements to adopt the provisions of Industry/
Technical Specification Task Force (TSTF) change TSTF-359, ``Increase
Flexibility in Mode Restraints.''
Date of issuance: June 25, 2004.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment Nos.: 281 and 265.
[[Page 43465]]
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 14, 2003 (68 FR
59217).
The supplemental letters dated December 15, 2003, and February 27,
2004, provided clarifying information that did not change the scope of
the original Federal Register notice or the original no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 25, 2004.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-255, Palisades Plant,
Van Buren County, Michigan
Date of application for amendment: June 3, 2003, as supplemented by
letters dated October 6, 2003, January 15, and February 13, 2004.
Brief description of amendment: The amendment revises the operating
license and technical specifications to increase the licensed rated
power by 1.4 percent from 2530 megawatts thermal (MWt) to 2565.4 MWt
using measurement uncertainty recapture.
Date of issuance: June 23, 2004.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 215.
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 8, 2003 (68 FR
40714).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 23, 2004.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of application for amendments: February 20, 2004.
Brief description of amendments: The amendments revised the
Technical Specification requirements for Shift Technical Advisor
coverage.
Date of issuance: June 28, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 132 and 111.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 2004 (69 FR
19574).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 28, 2004.
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: June 6, 2003, as supplemented by letter
dated February 24, 2004.
Brief description of amendments: The amendments revise the
Technical Specifications (TSs) adopting the TS Task Force (TSTF)
Standard TS Change Traveler TSTF-360, Revision 1, ``DC Electrical
Rewrite.'' Specifically, the amendments revise the TS 3.8.4, ``DC
Sources-Operating,'' TS 3.8.5, ``DC Sources-Shutdown,'' TS 3.8.6,
``Battery Cell Parameters,'' and TS 5.5.19, ``Battery Monitoring and
Maintenance Program.''
Date of issuance: July 1, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 113 and 113.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 8, 2003 (68 FR
40721). The February 24, 2004, supplemental letter provided clarifying
information that did not change the scope of the original Federal
Register notice or the original no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 1, 2004.
No significant hazards consideration comments received: No.
Dated in Rockville, Maryland, this 12th day of July 2004.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 04-16157 Filed 7-19-04; 8:45 am]
BILLING CODE 7590-01-P