[Federal Register Volume 68, Number 101 (Tuesday, May 27, 2003)]
[Notices]
[Pages 28843-28864]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-12973]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision 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, May 2, 2003, through May 15, 2003. The last
biweekly notice was published on May 13, 2003 (68 FR 25648).
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.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public
[[Page 28844]]
and State comments received before action is taken. Should the
Commission take this action, it will publish in the Federal Register a
notice of issuance and provide for opportunity for a hearing 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
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By June 26, 2003, 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 2.714, 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 by the above date, the
Commission or an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, 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 factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of 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. Petitioner must provide 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 to relief. A
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, 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 with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 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
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, 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 filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for
[[Page 28845]]
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].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: April 21, 2003.
Description of amendment request: The licensee proposed to revise
Sections 3.7 and 4.7, ``Auxiliary Electrical Power,'' of the Technical
Specifications (TSs) to make them generally consistent with Nuclear
Regulatory Commission (NRC) guidance set forth in NUREG-1433,
``Standard Technical Specifications, General Electric Plants, BWR
[Boiling Water Reactor]/4,'' Revision 2, and with the NRC-approved
industry guidance identified as Technical Specification Task Force
(TSTF) traveler TSTF-360, Revision 1. The amendment would also add a
new Section 6.8.5, ``Station Battery Monitoring and Maintenance
Program.'' The resulting Sections 3.7, 4.7, and 6.8.5 will be
explicitly applicable to station batteries B and C, both safety-related
subsystems, and their associated battery chargers. The proposed
amendment would revise requirements concerning surveillance,
monitoring, and maintenance of the subject batteries and chargers.
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 three standards of 10 CFR 50.92(c) and performed its own.
The NRC staff's analysis is presented below:
The first standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
increase in the probability or consequences of an accident previously
evaluated. The proposed changes, if approved by the NRC, will be made
in a manner such that conservatism is maintained through compliance
with applicable NRC regulations and guidance. No hardware design change
is involved with the proposed amendment, thus there will be no adverse
effect on the functional performance of any plant structure, system, or
component (SSC). Consequently, all SSCs will continue to perform their
design functions with no decrease in their capabilities to mitigate the
consequences of postulated accidents. Station battery surveillance,
monitoring, and maintenance were not previously factored into the
probability of accidents, nor were they factored into scenarios of
previously analyzed accidents. Consequently, the proposed revision to
Sections 3.7, 4.7, and 6.8.5 of the TSs will lead to no increase in the
consequences of an accident previously evaluated, and no increase of
the probability of an accident previously evaluated.
The second standard requires that operation of the unit 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 amendment is not the result of a hardware
design change, nor does it lead to the need for a hardware design
change. There is no change in the methods the unit is operated. As a
result, all SSCs will continue to perform as previously analyzed by the
licensee, and previously evaluated and accepted by the NRC staff.
Therefore, the proposed amendment will not create the possibility of a
new or different kind of accident from any previously evaluated.
The third standard requires that operation of the unit in
accordance with the proposed amendment will not involve a significant
reduction in a margin of safety. Since the licensee did not propose to
exceed or alter a design basis or safety limit, the proposed amendment
will not affect in any way the performance characteristics and intended
functions of any SSC. Therefore, the proposed amendment does not
involve a significant reduction in a margin of safety.
Based on the NRC staff's analysis, 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: Kevin P. Gallen, Morgan, Lewis & Bockius,
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Richard J. Laufer.
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: April 15, 2003.
Description of amendments request: The amendments would revise
Sections 2.2, ``SL [Safety Limits] Violations,'' for reporting such
violations to positions in the plant organization; 5.2.1, ``Onsite and
Offsite Organization,'' for the position responsible for overall safe
plant operation; and 5.5.1, ``Offsite Dose Calculation Manual (ODCM),''
for the position that approves changes to the ODCM, of the Technical
Specifications (TSs). The revisions would account for the elimination
of the positions of Vice President, Nuclear Production, and Director,
Site Chemistry, and the assignment of the responsibilities of these
positions in the above TS sections to other positions in the plant
organization. Also, there would be the format change of adding the
title of Section 2.2 near the top of TS page 2.0-2.
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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
These changes involve minor changes in the organization of
PVNGS. It is expected that the organizational changes will have a
positive effect on the conduct of plant operations and safety-
related work. Functions which are necessary to operate the facility
safely and in accordance with the operating licenses, remain in the
re-aligned organization and will not affect the safe operation of
the plant and continue to ensure proper control of administrative
activities. The Quality Assurance (QA) organization reporting
structure has not been affected by these changes allowing the QA
organization to maintain the required authority and organizational
freedom to identify quality problems; to initiate, recommend, or
provide solutions; and to verify implementation of solutions. The
proposed changes will not affect the operation of structures,
systems, [or] components, and will not reduce programmatic controls
such that plant safety would be affected. (The changes in the plant
organization are also not initiators of an accident.) Therefore, the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
[[Page 28846]]
The proposed changes will not affect the operation of
structures, systems, [or] components, and will not reduce
programmatic controls such that plant safety would be affected. The
changes in the organization will continue to provide necessary
oversight and control of administrative processes. [The changes in
the plant organization are also not initiators of an accident.]
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
These changes are administrative and will not diminish any
organizational or administrative controls currently in place. The
proposed changes will not affect the operation of structures,
systems, [or] components, and will not reduce programmatic controls
such that plant safety would be affected. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
Based on the above, APS concludes that the activities associated
with the proposed amendment(s) present no significant hazards
consideration under the standards set forth in 10 CFR 50.92
``Issuance of Amendment,'' (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.
Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona
Public Service Company, PO Box 52034, Mail Station 7636, Phoenix,
Arizona 85072-2024.
NRC Section Chief: Stephen Dembek.
Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert
Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, Maryland
Date of amendment request: May 1, 2003.
Description of amendment request: The proposed amendment would
increase the maximum enrichment limit of the fuel assemblies that can
be stored in the Unit 1 spent fuel pool by taking credit for soluble
boron in maintaining acceptable margins of subcriticality. 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 proposed change will increase the maximum enrichment limit
of the fuel assemblies that can be stored in the Unit 1 spent fuel
pool (SFP) by taking credit for soluble boron in maintaining
acceptable margins of subcriticality. The proposed change will
modify Technical Specification 4.3.1 ``Criticality'' and add
Technical Specification 3.7.16 ``Spent Fuel Pool Boron
Concentration.'' The postulated accidents for the SFP are basically
four types: (1) dropped fuel assembly on top of the storage rack,
(2) a misloading accident, (3) an abnormal location of a fuel
assembly, and (4) loss-of-normal cooling to the SFP.
There is no increase in the probability of a fuel assembly drop
accident in the SFP when considering the higher enriched fuel or the
presence of soluble boron in the SFP water. Dropping a fuel assembly
on top of the SFP storage racks is not credible at Calvert Cliffs
due to the design of the spent fuel handling machine and due to the
height of the SFP storage racks. The handling of the fuel assemblies
has always been performed in borated water and will not change as a
result of crediting soluble boron in the SFP criticality analysis.
The proposed change does not change the general design and
characteristics of the fuel assemblies. Therefore, the proposed
change does not increase the probability of a fuel assembly drop
accident.
There is no increase in the probability of the accidental
misloading of irradiated fuel assemblies into the SFP storage racks
when considering the higher enriched fuel or the presence of soluble
boron in the SFP water for criticality control. Fuel assembly
placement will continue to be controlled pursuant to approved fuel
handling procedures.
Due to the design of the SFP storage racks, an abnormal
placement of a fuel assembly into the SFP storage racks is not
possible. Also, the design of the SFP prevents an inadvertent
placement of a fuel assembly between the outer most storage cell and
the pool wall. The proposed change does not make any change to the
design of SFP. Therefore, there is no increase in the probability of
abnormal placement of a fuel assembly into the SFP storage racks.
The proposed change will not result in any changes to the SFP
cooling system, and the fuel assembly design and characteristics are
not changed by an increase in fuel enrichment. Therefore, there is
no increase in the probability of a loss of SFP cooling. Also, since
a high concentration of soluble boron has always been maintained in
the SFP water, there is no increase in the probability of the loss
of normal cooling to the SFP water considering the presence of
soluble boron in the pool water for criticality control.
There is no increase in the consequences of an accidental drop
or accidental misloading of a maximum enriched fuel assembly into
the SFP storage racks, because the criticality analysis demonstrates
that the pool will remain subcritical following either event, even
if the pool contains a boron concentration less than the proposed
Technical Specification limit. The proposed Technical Specification
limit will ensure that an adequate SFP boron concentration will be
maintained.
There is no increase in the consequences of a loss-of-normal SFP
cooling because the Technical Specification boron concentration
provides significant negative reactivity. Loss of the SFP water via
boiling will not result in a loss of soluble boron, since the
soluble boron is not volatile. Therefore, loss of spent fuel pool
cooling system without makeup flow is not a mechanism for boron
dilution. Even in the unlikely event that soluble boron in the SFP
is completely diluted via unborated makeup flow, a pool completely
filled with maximum enriched unburned assemblies will remain
subcritical by a design margin of k-effective not to exceed 0.986.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change will increase the maximum enrichment limit
of the fuel assemblies that can be stored in the Unit 1 SFP by
taking credit for soluble boron in maintaining acceptable margins of
subcriticality. Increasing the maximum enrichment limit does not
create a new type of criticality accident.
Soluble boron has been maintained in the SFP water and is
currently required by procedures. Therefore, crediting soluble boron
in the SFP criticality analysis will have no effect on normal pool
operation and maintenance. Crediting soluble boron will only result
in increased sampling to verify the boron concentration. This
increased sampling will not create the possibility of a new or
different kind of accident.
A dilution of the SFP soluble boron has always been a
possibility. However, the boron dilution event previously had no
consequences, since boron was not previously credited in the
accident analysis. The initiating events that were considered for
having the potential to cause dilution of the boron in the SFP to a
level below that credited in the criticality analyses fall into
three categories: dilution by flooding, dilution by loss-of-coolant
induced makeup, and dilution by loss-of-cooling system induced
makeup. The spent fuel pool dilution analysis demonstrates that a
dilution that could increase the rack k-effective greater than 0.95
is not a credible event. It is not credible that dilution could
occur for the required length of time without operator notice, since
this event would activate the high level alarm and initiate
Auxiliary Building flooding. In addition, in excess of 1,043,000
gallons of unborated water must be added to the SFP to reach the
minimum soluble boron concentration. This is more water volume than
is contained in both pretreated water storage tanks and also more
water volume than is contained in the demineralized water storage
tank and both condensate storage tanks combined. Even in the
unlikely event that soluble boron in the SFP is completely diluted,
the SFP will remain subcritical by a design margin of k-effective
will not exceed 0.986.
The proposed change will not result in any other change in the
plant configuration or equipment design. Therefore, the proposed
[[Page 28847]]
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The Technical Specification changes proposed by this license
amendment request will provide an adequate safety margin to ensure
that the stored fuel assembly array of maximum enriched fuel will
always remain subcritical. Those limits are based on a plant
specific criticality analysis performed for the Calvert Cliffs Unit
1 SFP, that include technically supported margins.
While the criticality analysis utilized credit for soluble
boron, the SFP rack k-effective will remain less than 0.986 with no
soluble boron with a 95 percent probability at a 95 percent
confidence level. This substantial reduction in the SFP soluble
boron concentration was evaluated and shown not to be credible.
Soluble boron is used to provide subcritical margin such that the
spent fuel pool k-effective is maintained less than or equal to
0.95. Since k-effective is less than or equal to 0.95, the current
margin of safety is maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposed to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
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: April 17, 2003.
Description of amendments request: The proposed amendment would (1)
make 19 specific changes to the Technical Specifications actions
currently requiring suspension of operations involving positive
reactivity additions, and (2) revise various notes precluding reduction
in boron concentration. The proposed changes follow the guidance of
Technical Specification Task Force (TSTF) Change Traveler 286, Revision
2.
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 those Technical
Specifications 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 operational difficulties. These controlled
activities could result in a slight change in the probability of an
event occurring as Reactor Coolant System (RCS) manipulations that
are 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 these conditions to allow for these
slight reactivity changes without significantly increasing the
probability of an accident previously evaluated.
The proposed change 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. In addition, small
temperature changes in the RCS impose reactivity changes by means of
the moderator temperature coefficient of reactivity. These small
changes are within the required shutdown margin, therefore, there is
no increase in the consequence of an accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different kind
of accident from any accident previously evaluated.
This proposed amendment allows for minor plant operational
adjustments 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.
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, 5, and 6 is preserved by the
calculated shutdown margin which prevents a return to criticality.
The proposed change will permit reductions in the discretionary
shutdown margin beyond the Technical Specification requirements.
However, the shutdown margin required by the Technical
Specifications is not changed. The proposed change only affects
Reactor Coolant System temperature and boron concentration above the
calculated shutdown margin. By not impacting the shutdown margin,
the margin of safety is not affected.
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
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Richard J. Laufer.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: February 13, 2003.
Description of amendment request: The proposed amendment would
allow the use of an alternative source term (AST) methodology in
accordance with 10 CFR 50.67 based on a reevaluation of the loss-of-
coolant accident (LOCA) design-basis accident (DBA). Using an approved
AST, the licensee has also proposed changes to increase the allowable
secondary containment bypass and main steam isolation valve (MSIV)
leakage limits and eliminate the MSIV leakage control system. The
licensee also proposed changes to the TS Bases.
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. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The implementation of AST assumptions has been evaluated in a
revision to the analysis of the Loss of Coolant Accident (LOCA) and
an update to the analysis of the Fuel Handling Accident (FHA).
Based upon the results of the analyses, it has been demonstrated
that, with the requested changes, the dose consequences of these
limiting Design Basis Accidents (DBAs) are within the regulatory
guidance provided by the NRC for use with the AST. This guidance is
presented in 10 CFR 50.67, Regulatory Guide 1.183 [``Alternative
Radiological Source Terms For Evaluating Design Basis Accidents At
Nuclear Power Reactors''], and Standard Review Plan (SRP) Section
15.0.1.
The requirements for MSIV [main steam isolation valve] Leakage
Control System operability for eliminating MSIV leakage to the
environment are being eliminated. This is acceptable because, with
the application of AST, this system is no longer credited in
mitigating the consequences of a LOCA or any other DBA.
The proposed changes also increase the limits on maximum
allowable leakage from
[[Page 28848]]
secondary containment bypass and main steam isolation valves, and on
unfiltered inleakage into the Control Room. This is acceptable due
to the new assumptions used in calculating Control Room and offsite
dose following the affected design basis accident using the AST
methodology.
The proposed changes do not affect the normal design or
operation of the facility before the accident; rather, once the
occurrence of an accident has been postulated, the new source term
is an input to evaluate the consequence. The radiological
consequences of the analyzed DBAs have been evaluated with
application of AST assumptions. The results conclude that the
radiological consequences remain within applicable regulatory
limits. Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The application of AST does not affect the design, functional
performance or normal operation of the facility. Similarly, it does
not affect the design or operation of any component in the facility
such that new equipment failure modes are created. Elimination of
the MSIV Leakage Control System cannot create a new accident because
it is used as a mitigation system to limit MSIV leakage after the
accident has occurred. Similarly, the use of Standby Liquid Control
System to buffer suppression pool pH to prevent iodine reevolution
is another mitigation function credited after the accident has
occurred and; therefore, cannot create a new accident.
As such the proposed changes will not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
This proposed license amendment involves changes from the
original source term developed in accordance with Technical
Information Document (TID) 14844 to a new AST, as described in
Regulatory Guide 1.183. The results of the DBA analyses and the
requested Technical Specification changes, are subject to revised
acceptance criteria. The analyses have been performed using
conservative methodologies.
Safety margins and analytical conservatisms have been evaluated
and have been found acceptable. The analyzed events have been
carefully selected and margin has been retained to ensure that the
analysis adequately bounds postulated event scenario. The dose
consequences of these limiting events are within the acceptance
criteria presented in 10 CFR 50.67, Regulatory Guide 1.183 and SRP
Section 15.0.1.
The margin of safety is that provided by meeting the applicable
regulatory limits. The effect of relaxation of these design and
Technical Specification requirements has been analyzed and doses
resulting from the design basis accidents have been found to remain
within the regulatory limits. The changes continue to ensure that
the doses at the exclusion area and low population zone boundaries,
as well as the control room, are within the corresponding regulatory
limits.
Therefore, operation of Fermi 2 in accordance with the proposed
changes 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
amendment request involves no significant hazards consideration.
Attorney for licensee: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
NRC Section Chief: L. Raghavan.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: February 13, 2003.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) Section 5.5.10, ``Technical
Specification (TS) Bases Control Program,'' to be consistent with
changes made to 10 CFR 50.59, which were published in the Federal
Register on October 4, 1999 (64 FR 53582), and which became effective
March 13, 2001.
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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change deletes the reference to ``unreviewed safety
question'' as defined in 10 CFR 50.59. Deletion of the definition of
``unreviewed safety question'' was approved by the NRC with the
revision of 10 CFR 50.59. This change is administrative in nature.
Consequently, the probability of an accident previously evaluated is
not significantly increased. Changes to the TS Bases are still
evaluated in accordance with 10 CFR 50.59. As a result, the
probability or consequences of any accident previously evaluated are
not significantly affected. There is no increase in the radiological
dose at the site boundary for any previously evaluated accident.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not involve a physical alteration of
the plant (i.e., no new or different types of equipment will be
installed) or a change to the methods governing normal plant
operation. These changes are considered administrative in nature and
do not modify, add, delete, or relocate any technical requirements
in the TS. Therefore, this change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The change does not involve a significant reduction in the
margin of safety.
The proposed change will not reduce a margin of safety because
it has no direct effect on any of the safety analysis assumptions.
Changes to the TS Bases that result in meeting the criteria in
paragraph 10 CFR 50.59(c)(2) continue to require NRC approval
pursuant to 10 CFR 50.59. This change is administrative in nature
based on the revision to 10 CFR 50.59. Therefore, the proposed
change does 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: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
NRC Section Chief: L. Raghavan.
Detroit Edison Company (DECo), Docket No. 50-341, Fermi 2, Monroe
County, Michigan.
Date of amendment request: March 31, 2003.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Surveillance Requirement (SR)
3.7.3.6 associated with the verification of control room emergency
filtration (CREF) system duct work unfiltered inleakage. This amendment
request supercedes DECo's previous amendment request dated September
26, 2002, in its entirety. The September 26, 2002, amendment request
was previously noticed in the Federal Register on November 26, 2002 (67
FR 70765).
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. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
This license amendment proposes an alternative test for
performing the (Control Room Emergency Filtration) CREF system
surveillance associated with measuring the
[[Page 28849]]
Control Room Envelope (CRE) unfiltered inleakage. The CREF system
provides a configuration for mitigating radiological consequences of
accidents; however, it does not involve the initiation of any
previously analyzed accident. Similarly, the implementation of
compensatory measures to address the failure of the surveillance to
meet the design basis unfiltered inleakage limits is required to
mitigate the consequences of a radiological release. Therefore, the
proposed changes cannot increase the probability of any previously
evaluated accident.
The CREF system provides a radiologically controlled environment
from which the plant can be safely operated following a radiological
accident. Design basis accident analyses conclude that radiological
consequences are within the regulatory acceptance criteria. The
current TS surveillance (SR 3.7.3.6) measures inleakage from four
sections of CREF system duct work outside the CRE that are at
negative pressure during accident conditions. The proposed Tracer
Gas test provides a measurement of CRE inleakage from all potential
sources including the four sections of duct work. Measuring the CRE
inleakage using Tracer Gas testing has no effect on the CREF system
function. The results of Tracer Gas testing will be evaluated
against the assumptions in the approved Alternative Source Term
(AST) design basis accident analyses and compensatory measures will
be implemented, as necessary, to ensure compliance with 10 CFR
50.67. If compliance with 10 CFR 50.67 cannot be demonstrated or if
compensatory measures have been in place for more than 18 months, a
conservative plant shutdown will be required to minimize risk.
Therefore, the proposed changes do not significantly increase the
radiological consequences of any previously evaluated accident.
Based on the above, the proposed changes do not significantly
increase the probability or consequences of any accident previously
evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not alter the design function or
operation of the system involved. The CREF system will still provide
protection to control room occupants in case of a significant
radioactive release. The revised TS surveillance requirements
provide an alternative test method that has been widely accepted for
the measurement of CRE unfiltered inleakage. The proposed changes do
not introduce any new modes of plant or CREF system operation.
Therefore, the proposed changes do not create the potential for a
new or different kind of accident from any accident previously
evaluated.
3. The changes do not involve a significant reduction in the
margin of safety.
The proposed changes to the Fermi 2 TS surveillance requirements
do not affect the radiological release from a design basis accident
nor the postulated dose to the control room occupants as a result of
the accident. The alternate surveillance test requirements provide
an acceptable approach for the measurement of CRE inleakage. Safety
margins and analytical conservatisms are included in the analyses to
ensure that all postulated event scenarios are bounded. The proposed
TS requirements continue to ensure that the radiological
consequences at the control room are below the corresponding
regulatory guidelines and that compliance with 10 CFR 50.67 and GDC
(General Design Criterion)-19 is not affected. Therefore, the
proposed changes will not result in 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: Peter Marquardt, Legal Department, 688 WCB,
Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-1279.
NRC Section Chief: L. Raghavan.
Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone
Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: April 7, 2003
Description of amendment request: The proposed amendment would move
selected Technical Specification (TS) parameters to the Core Operating
Limits Reports (COLR). Specifically, the changes proposed affect TSs
2.2, ``Limiting Safety System Settings, Table 2.2-1;'' 3/4.1.1.1.1,
``Reactivity Control Systems, Boration Control, SHUTDOWN MARGIN--Modes
3, 4, and 5 Loops Filled;'' 3/4.1.1.2, ``Reactivity Control Systems,
SHUTDOWN MARGIN--Cold Shutdown--Loops Not Filled;'' 3/4.2.5, ``Power
Distribution Limits, DNB Parameters;'' 3/4.3.5, ``Instrumentation,
SHUTDOWN MARGIN Monitor;''
3/4.9.1.1, ``Refueling Operations, Boron Concentration;'' Section
6.9.1.6.a, ``Core Operating Limits Report, Core Operating Limits;'' and
Section 6.9.1.6.b, ``Core Operating Limits Report, The Analytical
Methods Used to Determine the Core Operating Limits,'' and the
corresponding pages and Bases sections.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR), Sec. 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 relocation of cycle-specific core operating limits from the
technical specifications to the COLR has no influence or impact on
the probability or consequences of a Design Basis Accident.
Adherence to the COLR and methodologies acceptable for establishing
COLR parameters continues to be controlled by Technical
Specifications. The proposed amendment still requires exactly the
same actions to be taken when or if limits are exceeded. Each
accident analysis addressed in the Final Safety Analysis Report
(FSAR) will be examined with respect to the changes in cycle-
dependent parameters, which are obtained from application of the
Nuclear Regulatory Commission (NRC) approved reload design
methodologies, to ensure that the transient evaluation of new core
designs are bounded by previously accepted analysis. This
examination, which will be performed in accordance with the
requirements of 10 CFR 50.59, ensures that future designs will not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed change to add new document references to Technical
Specification Sections 6.9.1.6.b.16 and 6.9.1.6.b.17 are required to
identify the most recent methodology to be used in the Millstone
Unit No. 3 Small Break Loss of Coolant Accident (SBLOCA) analysis.
Section 6.9.1.6.b.18 is added to describe NRC approved Overpower DT
and Overtemperature DT trip function methodology. The use of these
methodologies demonstrates that the acceptance criteria for SBLOCA
events and Overpower DT and Overtemperature DT are met. This change
has no impact on plant equipment operation. Since these changes only
affect the method of analysis, they cannot affect the likelihood or
consequences of accidents. Therefore, these changes will not
increase the probability or consequences of an accident previously
evaluated.
Deleting the revision number and the date from the documents
contained in Technical Specification Section 6.9.1.6.b.1 and in
Technical Specification Sections 6.9.1.6.b.4 through 6.9.1.6.b.10
has no impact on the actual analytical methods used to determine the
core operating limits, nor does it affect the likelihood or
consequences of accidents. Therefore, this change will not increase
the probability or consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
As stated earlier, the relocation of the cycle-specific
variables to the COLR, adding new document references and deleting
the revision number and the date in Technical Specification Section
6.9.1.6.b have no influence or impact, nor does it contribute in any
way to the probability or consequences of an accident. No safety
related equipment, safety function, or plant operations will be
altered as a result of this proposed change. The cycle specific
variables are calculated using NRC-approved methods and submitted to
the NRC to allow the Staff to continue to trend the values of these
limits. The Technical Specifications will continue to require
operation within the required core operating limits and appropriate
actions will be taken when or if limits are exceeded. Therefore the
proposed amendment does not in any way create the possibility of a
new or
[[Page 28850]]
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes have no impact on plant equipment
operation. The proposed changes do not revise any setpoints assumed
in the analyses and do not affect the acceptance criteria for SBLOCA
analyses. Therefore, the proposed changes will not result in a
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
NRC Section Chief: James W. Clifford.
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: April 10, 2003.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) for the low temperature
overpressure protection system. Currently, TS Surveillance Requirement
(SR) 3.4.12.5 requires performance of a channel functional test for the
power-operated relief valve within 12 hours of decreasing reactor
coolant system (RCS) temperature to <= 325 [deg]F and every 31 days
thereafter. The proposed amendments would revise TS SR 3.4.12.5 to
allow the first performance of this surveillance to be within 31 days
prior to decreasing RCS temperature to <= 325 [deg]F. The proposed
amendments also would revise the frequency of the channel calibration
in TS SR 3.4.12.7 from 18 months to 6 months.
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:
Pursuant to 10 CFR 50.91, Duke Power Company (Duke) has made the
determination that this amendment request involves a No Significant
Hazards Consideration by applying the standards established by the
NRC regulations in 10 CFR 50.92. This ensures 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:
This is a revision to the Technical Specification (TS)
surveillance requirement (SR) for performing the channel functional
test (CFT) for the pressurizer [power] [..] operated relief valve
(PORV). As such, changing the requirement to perform the first CFT
before entering the Low Temperature Overpressure Protection (LTOP)
region, rather than after LTOP is required, eliminates removing the
PORV from service, in the mode of applicability for the performance
of the CFT. This change will decrease the probability of a low
temperature overpressurization of the reactor vessel, thereby
increasing safety and reducing risk, by maintain(ing) both trains
(active and passive) of the LTOP System operable. The change to the
frequency for performance of SR 3.4.12.7 is being done to ensure the
calibration is performed in a time frame supported by current
analysis. The method of test is not changed, only the frequency.
This reduction in frequency will not significantly increase the
probability or consequences of any accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated:
This revision will not impact the LTOP evaluation analysis. The
timeframe to perform the CFT for the PORV will not change the
operation of the PORV or its function during accident conditions. No
new or different accidents result from performing the CFT prior to
entering LTOP conditions. The revision to SR 3.4.12.7 only changes
the frequency of the testing. The method of test is not changed.
This change has no effect on the possibility of a new or different
kind of accident.
(3) Involve a significant reduction in a margin of safety:
The proposed revision will perform the CFT within 31 days prior
to entering LTOP conditions, rather than performing the test once
LTOP conditions are entered. This allows the CFT, which causes the
PORV to be inoperable for a short period of time, to be performed
prior to reaching the plant conditions where the PORV is relied upon
for LTOP. Performing the CFT within 31 days prior to decreasing RCS
temperature to < 325 [deg][F], rather than after entering these
conditions, will not change the margin of safety. Oconee
calculations show that a recalibration interval of 6 months for the
Reactor Coolant System (RCS) low range pressure instrumentation
results in a single-sided 95/95 probability confidence limit of 9.4
psig. This result is bounded by the instrument uncertainty assumed
in the LTOP evaluation analysis. The frequency change for SR
3.4.12.7 from 18 months to 6 months does not affect the method of
test performance. It only decreases the allowed time between
performances to reflect current Oconee analysis. This will not
significantly reduce 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. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC 20005.
NRC Section Chief: John A. Nakoski.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: July 5, 2002, as supplemented August 13,
2002.
Description of amendment request: The proposed amendment would
relocate portions of Technical Specification (TS) 3/4.6.B, ``Primary
System Boundary--Coolant Chemistry,'' from the TSs to the Updated Final
Safety Analysis Report (UFSAR). The portions of the TS that would be
relocated to the UFSAR are the reactor coolant chemistry requirements
for conductivity and chloride concentration. Specifically, TSs 3/
4.6.B.2, 3/4.6.B.3, and 3.6.B.4 would be relocated to the UFSAR.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (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 is administrative in nature
and does not involve the modification of any plant equipment or
affect basic plant operation. Conductivity and chloride limits are
not assumed to be an initiator of any analyzed event, nor are these
limits assumed in the mitigation of consequences of accidents.
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 does not involve any physical
alteration of plant equipment and does not change the method by
which any safety-related system performs its function. As such, no
new or different types of equipment will be installed, and the basic
operation of installed equipment is unchanged. The methods governing
plant operation and testing remain consistent with current safety
analysis assumptions. Therefore, the proposed 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 represents the relocation of
current Technical
[[Page 28851]]
Specification requirements to the UFSAR, based on regulatory
guidance and previously approved changes for other stations. The
proposed change is administrative in nature, does not negate any
existing requirement, and does not adversely affect existing plant
safety margins or the reliability of the equipment assumed to
operate in the safety analysis. As such, there are no changes being
made to safety analysis assumptions, safety limits or safety system
settings that would adversely affect plant safety as a result of the
proposed change. Margins of safety are unaffected by requirements
that are retained, but relocated from the Technical Specifications
to the UFSAR. 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: J.M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts, 02360-5599.
NRC Section Chief: James W. Clifford.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: May 1, 2003.
Description of amendment request: The proposed amendment would
modify the surveillance testing requirements for the containment spray
system (CSS) by deleting the requirement to verify the position of
valves that are locked, sealed, or otherwise secured in their correct
position and replacing the quantitative allowable pump degradation
value with a requirement to verify the pumps perform in accordance with
the Inservice Testing Program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Analyzed events are assumed to be initiated by the failure of
plant structures, systems, or components. Altering the surveillance
requirements for the CSS does not increase the probability that a
failure leading to an analyzed event will occur. The CSS components
are passive until an actuation signal is generated. This change does
not increase the failure probability of the CSS components.
Therefore, the probability of occurrence for a previously analyzed
accident is not significantly increased.
The CSS is primarily designed to mitigate the consequences of a
loss of coolant accident (LOCA) or main steam line break (MSLB)
accident. The proposed change does not affect any of the assumptions
used in the deterministic LOCA or MSLB analyses. Hence the
consequences of accidents previously evaluated do not change.
Therefore, the change associated with modifying the CSS
surveillance requirements does not involve an increase in the
probability or consequences of any 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 does not change the design or configuration
of the plant. No new equipment is introduced, nor will any installed
equipment be operated in a new or different manner. No changes are
proposed to the plant's operating parameters or setpoints at which
protective or mitigative actions are initiated. Additionally, no
substantive changes are proposed to the procedures which ensure the
plant remains within analyzed limits or the procedures relied upon
to respond to off-normal events. As such, no new failure modes are
being introduced. The proposed change does not alter assumptions
made in the safety analysis or licensing basis.
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 associated with modifying the surveillance
requirements for the CSS does not affect the limiting conditions for
operation used in the deterministic analysis to establish the margin
of safety. The margin of safety is established through equipment
design, operating parameters, and the setpoints at which automatic
actions are initiated. None of these are adversely impacted by the
proposed change. Sufficient equipment remains available to actuate
upon demand for the purpose of mitigating a transient event.
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: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: March 11, 2003.
Description of amendment request: The proposed amendment will
revise and relocate Surveillance Requirement (SR) 4.0.5 and SR 4.4.9 to
the administrative section of the Technical Specifications (TS) under
sections 6.5.8 and 6.5.7, respectively. The proposed amendment will
also relocate TS 3.4.9, ``Reactor Coolant System Structural Integrity''
and its Bases to the Waterford Steam Electric Station, Unit 3
(Waterford 3) Technical Requirements Manual (TRM). Additionally, the
proposed amendment extends the Waterford 3 flywheel volumetric
examination interval to ten years.
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 to relocate SR 4.0.5 to the administrative
section of the TSs, including modifications to the wording to make
it consistent with NUREG-1432, will not reduce the current testing
and inspection requirements. The performance of a code (American
Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel
Code) inservice test is not an accident initiator. Verbally issuing
relief to the ASME Code by the NRC (Nuclear Regulatory Commission)
staff in lieu of written relief does not reduce assurance of the
health and safety of the public since the NRC staff still reviews
the basis for the relief on its technical merit and the NRC staff
still obtains management approval prior to granting the relief.
Inspections of the reactor coolant pump (RCP) flywheels are
conducted to detect a flaw in the flywheel prior to it becoming a
missile that could damage other portions of the facility. The
fracture mechanics analyses conducted as part of NRC approved
Topical Report SIR-94-080-A, Rev(ision) 1 shows that a
conservatively sized pre-existing crack will not grow to a flaw size
necessary to create flywheel missiles within the current or extended
life of the facility thus the flywheel will remain intact and
perform its function to reduce the rate of decay of coolant flow
during a postulated loss of power to the RCP motor. This analysis
conservatively assumes minimum material properties, maximum flywheel
speed, location of the flaw in the highest stress area, and a number
of startup and shutdown cycles higher than expected. Since a
conservative flaw in the RCP flywheels will not grow to the
allowable flaw size under large break LOCA (loss-of-coolant
[[Page 28852]]
accident) conditions over the life of the plant, reducing the
inspection frequency of the flywheels will not significantly
increase the probability or consequences of an accident previously
evaluated.
The change to move the surveillance requirements for the RCP
flywheels to the programs section of the technical specifications is
administrative and has no impact on probability or consequences of
an accident.
The change to move TS 3.4.9 to the Waterford 3 TRM will have no
adverse effect on plant operation or the availability or operation
of any accident mitigation equipment. Changes to the TRM are
controlled in accordance with 10 CFR 50.59. Therefore, moving TS
3.4.9 to the Waterford 3 TRM will not adversely impact [as] an
accident initiator and can not cause an accident.
Therefore, the proposed 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 proposed changes will not alter the plant configuration (no
new or different type of equipment will be installed) or require any
new or unusual operator actions. They do not alter the way any
structure, system, or component functions and do not alter the
manner in which the plant is operated. These changes do not
introduce any new failure modes.
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 testing and inspection requirements contained in TS 4.0.5
are governed by 10 CFR 50.55a, ``Codes and Standards.'' The 10 CFR
requirements to perform the ASME code testing and inspections will
not be reduced by the proposed change. The inspections and tests
will continue to be performed as they are currently. The proposed
change has no impact on plant equipment operation.
The fracture mechanics analysis conducted in support of
extending the RCP flywheel volumetric examination interval from
three years to ten years shows that significant conservatism has
been used for calculating the allowable flaw size, critical flaw
size, and crack growth rate in the RCP flywheels. These include
minimum material properties, maximum flywheel accident speed,
location of the flaw in the highest stress area, and a number of
startup/shutdown cycles eight times greater than expected. Since a
postulated flaw in a Waterford 3 flywheel will not grow to the
allowable flaw size under normal operating conditions or to the
critical flaw size under loss of coolant accident conditions over
the life of the plant, reducing the examination requirements for the
detection of such cracks over the life of the plant will not involve
a significant reduction in the margin of safety. The proposed change
has no impact on plant equipment operation.
The change to move the surveillance requirements for the RCP
flywheels to the programs section of the technical specifications is
administrative and has no impact on plant operation.
Relocation of TS 3.4.9 to the TRM does not imply any reduction
in its importance in ensuring that the structural integrity and
operational readiness of ASME Code Class 1, 2, and 3 components will
be maintained at an acceptable level throughout the life of the
plant. The proposed change has no impact on plant operation.
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: N.S. Reynolds, Esquire, Winston & Strawn,
1400 L Street NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: March 31, 2003.
Description of amendment request: The proposed amendments would
revise Appendix A, Technical Specifications (TS), of Facility Operating
License Nos. NPF-11 and NPF-18. Specifically, the proposed change will
modify TS 5.7, ``High Radiation Area,'' by incorporating the wording
and requirements from NUREG-1434, ``Standard Technical Specifications
General Electric Plants, BWR/6,'' Revision 2, dated June 2001.
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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change will modify LaSalle County Station (LSCS) TS
5.7, ``High Radiation Area,'' by incorporating into the TS the
corresponding wording and requirements from NUREG-1434, ``Standard
Technical Specifications General Electric Plants, BWR/6,'' Revision
2, dated June 2001. TS 5.7 establishes the administrative controls
on entry into high radiation areas. High radiation area
administrative controls are not a precursor to accidents previously
evaluated. Thus, the proposed change does not have any effect on the
probability of an accident previously evaluated.
The proposed change in administrative controls on entry into a
radiation area does not affect the ability of LSCS to successfully
respond to previously evaluated accidents and does not affect
radiological assumptions used in the evaluations. Thus, the
radiological consequences of any accident previously evaluated are
not increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change does not affect the control parameters
governing unit operation or the response of plant equipment to
transient conditions. The proposed change does not introduce any new
equipment, modes of system operation or failure mechanisms.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change incorporates corresponding wording and
requirements from NUREG-1434, into the LSCS TS. The LSCS evaluation
of the proposed change concluded the following:
[sbull] Both the proposed and current TS 5.7.1 describe the
requirements for access into areas that have radiation levels that
exceed 100 mrem/hr but are less than or equal to 1000 mrem/hr. The
proposed and current TS 5.7.1 are considered to have equivalent
level access controls as both contain the need to provide a
barricade, conspicuously post the area and issue an RWP to control
entrance to the area.
[sbull] Proposed TS 5.7.2 and current TS 5.7.4 describe the
requirements for access into areas that have radiation levels that
exceed 1000 mrem/hr. Proposed TS 5.7.2 and current TS 5.7.4 are
considered to have equivalent level access controls as both require
these areas to be locked. For those areas where locking is not
practical, proposed TS 5.7.2 and current TS 5.7.4 both require the
area to be barricaded, conspicuously posted, and have an activated
flashing light.
[sbull] The proposed change includes the deletion of the use of
computer controlled doors in current TS 5.7.2. This description is
being removed as computer controlled doors are no longer utilized at
LSCS. Rather, manual locking mechanisms are used on doors providing
an equivalent level of control.
[sbull] Current TS 5.7.4 also discusses ``high-high'' radiation
areas. The term ``high-high'' radiation area is a legacy term that
is being deleted from the proposed TS. This is an administrative
change only to remove an outdated term.
Therefore, LSCS has determined that the proposed change provides
an equivalent level of protection as that currently provided.
[[Page 28853]]
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
Based upon the above, Exelon Generation Company 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
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. Edward J. Cullen, Deputy General
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
NRC Section Chief: Anthony J. Mendiola.
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: March 11, 2003.
Description of amendment request: The proposed amendment revises
the BVPS-1 and 2 Technical Specifications (TSs) to apply the
Westinghouse best-estimate large break loss-of-coolant accident (LOCA)
methodology to BVPS-1 and 2. The request is contingent upon Nuclear
Regulatory Commission (NRC) approval of the licensee's amendment
request for conversion of the BVPS-1 and 2 containments from sub-
atmospheric to atmospheric which had previously been requested by
letter dated June 5, 2002, and which is currently under NRC staff
review.
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?
No. No physical changes are required as a result of implementing
best-estimate loss of cooling accident (LOCA) methodology and
associated technical specification changes. The plant conditions
assumed in the analysis are bounded by the design conditions for all
equipment in the plant. Therefore, there will be no increase in the
probability of a loss of cooling accident. The consequences of a LOCA
are not being increased, since it is shown that the emergency core
cooling system is designed so that its calculated cooling performance
conforms to the criteria contained in 10 CFR 50.46, Paragraph b. No
other accident is potentially affected by this change.
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 previously evaluated?
No. There are no physical changes being made to the plants. No new
modes of plant operation are being introduced. The parameters assumed
in the analysis are within the design limits of the existing plant
equipment. All plant systems will perform as designed during the
response to a potential 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?
No. It has been shown that the methodology used in the analysis
would more realistically describe the expected behavior of plant
systems during a postulated loss of coolant accident. Uncertainties
have been accounted for as required by 10 CFR 50.46. A sufficient
number of loss of coolant accidents with different break sizes,
different locations and other variations in properties are analyzed to
provide assurance that the most severe postulated loss of coolant
accidents are calculated. It has been shown by analysis that there is a
high level of probability that all criteria contained in 10 CFR 50.46,
Paragraph b are met.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
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.
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit 2, Berrien County, Michigan
Date of amendment request: March 27, 2003.
Description of amendment request: The proposed amendment would
amend Unit 2 Technical Specification (TS) Table 3.3-4 and the P-11
setpoint in the Engineered Safety Features Interlock Table as follows:
1. Revise the low pressurizer pressure safety injection (SI) trip
setpoint from its current value of greater than or equal to 1900 pounds
per square inch gauge (psig), to greater than or equal to 1815 psig.
2. Revise the low pressurizer pressure SI allowable value from
greater than or equal to 1890 psig, to greater than or equal to 1805
psig.
3. Revise the P-11 setpoint from its current value of greater than
or equal to 2010 psig, to greater than or equal to 1915 psig.
4. Make format changes to the affected TS pages that improve
appearance but do not affect any requirements.
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 the consequences of an accident previously
evaluated?
Response: No.
[Indiana Michigan Power Company] (I&M) proposes changing the low
pressurizer pressure SI trip setpoint, the low pressurizer pressure
SI allowable value, the P-11 setpoint, and the format of the
associated pages. Neither the change to the low pressurizer pressure
SI trip setpoint value and the SI allowable value nor the change to
the P-11 setpoint value alter any safety-related components or the
means of accomplishing a safety-related function. The change in the
values is supported by analyses that demonstrate that applicable
acceptance criteria are met when SI is initiated at 1700 psig for a
(loss-of-coolant accident) LOCA, a main steam system
depressurization event, and a feedwater line break. Because the
acceptance criteria are met, there is no significant increase in the
consequences of an accident. The format changes are intended to
improve readability and appearance, and do not alter any
requirements. Thus, neither the probability of an accident nor the
consequences of an accident are significantly increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
[[Page 28854]]
I&M proposes changing the low pressurizer pressure SI trip
setpoint, the low pressurizer pressure SI allowable value, the P-11
setpoint, and the format of the associated pages. Neither the change
to the low pressurizer pressure SI trip setpoint value and the SI
allowable value nor the change to the P-11 setpoint value involve
changing the design function of any component, and a change in any
of the values cannot initiate an accident. The format changes are
intended to improve readability and appearance, and do not alter any
requirements. Thus, no new accident initiators are introduced, and
the possibility of a new or different kind of accident is not
created.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
I&M proposes changing the low pressurizer pressure SI trip
setpoint, the low pressurizer pressure SI allowable value, the P-11
setpoint, and the format of the associate pages. The low pressurizer
pressure instrument is credited for activating the engineered safety
features in the event of a LOCA, a main steam system
depressurization event, or a feedwater line break. The low
pressurizer pressure SI trip setpoint value and the low pressurizer
pressure SI allowable value have been selected to insure that the
engineered safety features will be activated as assumed in the
safety analysis. Present margins continue to be maintained because
the applicable accident analyses criteria continue to be met. No
margins of safety are associated with the P-11 setpoint value. The
format changes are intended to improve readability and appearance,
and do not alter any requirements. Thus, there is no significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: L. Raghavan.
Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee
Atomic Power Station, Lincoln County, Maine
Date of amendment request: April 24, 2003.
Description of amendment request: Eliminate the requirement for
continuous Control Room manning when fuel is stored in the fuel storage
pool.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Involve a significant increase in the probability or consequence of
an accident previously evaluated.
Response: No.
The Defueled Safety Analysis (DSAR) identifies five categories
of events: spent fuel criticality accidents, a fuel handling
accident, a spent fuel cask drop, spent fuel pool accidents, and a
low level waste release incident. There are no active controls in
the control room that are required to respond to these events.
Actions to mitigate the consequences of these events are taken
outside the control room. Emergency response is not adversely
affected by this proposed change because the control room is still
available to the emergency response team and communications
capability and timeliness will not be affected. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The configuration, operations and accident response of the
systems, structures or components that support safe storage of the
spent fuel are unchanged by the proposed change to the technical
specifications. Current site surveillance requirements ensure
frequent and adequate monitoring of system and component
functionality. Systems in the Spent Fuel Pool Island will continue
to be operated in accordance with current design requirements and no
new components or system interactions have been identified. No new
accident scenarios, failure mechanisms or limiting single failures
are introduced as a result of the proposed change. The proposed
technical specification change does not have an adverse affect on
any system related to safe storage of spent fuel. Therefore, the
proposed technical specifications change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
All design basis accident acceptance criteria will continue to
be met. The margin of safety relative to the cooling of the spent
fuel is unaffected by the proposed changes as the spent fuel pool
parameters will continue to be monitored at the same frequency as
assumed in the accident analysis. The ability of the shift crew to
respond to abnormal or accident conditions is unaffected by the
proposed change since all controls are located in or near the fuel
building and any necessary communications will be handled by the on-
shift staff and/or DERO. Therefore, it is concluded that the
proposed TS change does not involve a significant reduction in the
margin of safety
Based on the above, Maine Yankee 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
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 amendment involves no significant hazards consideration.
Attorney for licensee: Joe Fay, Esquire, Maine Yankee Atomic Power
Company, 321 Old Ferry Road, Wiscasset, Maine 04578.
NRC Section Chief: Claudia M. Craig.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: October 8, 2002.
Description of amendment request: The license amendment request
proposes to change the title of (a) Shift Supervisor to Shift Manager,
(b) ``Plant Manager'' to ``plant manager,'' (c) ``Vice President--
Nuclear'' to ``corporate officer with direct responsibility for the
plant,'' (d) ``Radiological Manager'' to ``radiological manager,'' (e)
``Operations Supervisor'' to ``operations supervisor'' and (f) ``Shift
Radiological Protection/Chemistry Technician'' to ``radiation
protection technician.'' This proposal includes an Updated Safety
Analysis Report (USAR) reference correction resulting from the USAR
Rebaseline Project and a correction to the title ``Shift Technical
Advisor'' to ``Shift Technical Engineer'' in Technical Specification
(TS) Section 5.3.1 so as to be consistent with the title used in TS
Section 5.2.2.f. These changes do not eliminate any of the
qualifications, responsibilities, or requirements for these positions.
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?
The title of Shift Manager better conveys the appropriate level
of responsibility and authority required of the position. The use of
generic personnel titles and correction of the USAR reference are
strictly administrative. The qualifications, training, duties and
experience required of the individuals remain unchanged. The USAR
section to be referenced is physically the same section that was
referenced before the USAR renumbering. The requested changes do not
[[Page 28855]]
involve any change to the design basis of the plant or any
structure, system, or component. 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 accident previously evaluated?
There will be no physical alterations to the plant
configuration. No changes in operating mode or limits are proposed.
The qualifications, training, duties and experience required of the
individuals remain unchanged. The USAR section to be referenced is
physically the same section that was referenced before the USAR
renumbering. Therefore, these proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
The proposed change in titles and USAR reference are strictly
administrative. The qualifications, training, duties and experience
required of the individuals remain unchanged. The USAR section to be
referenced is physically the same section that was referenced before
the USAR renumbering. The proposed changes do not change any license
condition or Technical Specifications safety limit or limiting
condition for operation. The changes do not involve modification of
the design or operation of any plant system involved with
controlling the release of radioactivity to the environment.
Therefore, these changes do not involve a significant reduction in a
margin of safety.
Based on the above, Nebraska Public Power District 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
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 R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy
Center, Linn County, Iowa
Date of amendment request: May 2, 2003.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TSs) by replacing the existing
Reactor Coolant System (RCS) pressure and temperature (P/T) limit
curves for in-service leakage and hydrostatic testing, non-nuclear
heatup and cooldown, and criticality (Figure 3.4.9-1, ``Pressure Versus
Minimum Temperature Valid to Thirty-two Full Power Years, per Appendix
G of 10 CFR 50'') with new, updated P/T limits curves. The replacement
curves were generated using an NRC-approved methodology (General
Electric Report NEDC-32983P) for determining the neutron fluence on the
Reactor Pressure Vessel (RPV) and extends the RPV beltline region to
encompass a new limiting component, the recirculation inlet nozzle. The
change to Figure 3.4.9-1 would also delete the existing notation that
states: ``(Interim Approval Until September 1, 2003).''
The licensee's application for amendment dated May 2, 2003,
supersedes and withdraws a previous application dated February 28,
2003, for which the NRC has published a notice of consideration of
issuance of amendment, proposed no significant hazards consideration
determination, and opportunity for hearing in the Federal Register (68
FR 12954, dated March 18, 2003).
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) The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The P/T limits are not derived from Design Basis Accident (DBA)
analyses. They are prescribed by the (American Society of Mechanical
Engineers Boiler and Pressure Vessel Code) ASME Code and 10 CFR 50
Appendix G and H and associated guidance documents, such as
Regulatory Guide 1.99, Rev. 2, as restrictions on normal operation
to avoid encountering pressure, temperature, and temperature rate of
change conditions that might cause undetected flaws to propagate and
cause non-ductile failure of the reactor coolant pressure boundary.
Thus, they ensure that an accident precursor is not likely. Hence,
they are included in the TS as satisfying Criterion 2 of 10 CFR
50.36(c)(2)(ii). The revision of the numerical value of these
limits, i.e., new curves, using an NRC-approved methodology, does
not change the existing regulatory requirements, upon which the
curves are based. Thus, this revision will not increase the
probability of any accident previously evaluated.
The proposed change does not alter the design assumptions,
conditions, or configuration of the facility or the manner in which
the facility is operated or maintained. The proposed changes will
not affect any other System, Structure or Component (SSC) designed
for the mitigation of previously analyzed events. The proposed
change does not affect the source term, containment isolation, or
radiological release assumptions used in evaluating the radiological
consequences of any accident previously evaluated. Thus, the
proposed revision of the existing numerical values with the updated
figure for the RCS P/T limits, which are based upon an NRC-approved
methodology for calculating the neutron fluence on the RPV and new
limiting component, will not increase the consequences of any
previously evaluated accident.
(2) The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve a physical alteration of the
plant (i.e., no new or different type of equipment will be
installed) or a change in the processes governing normal plant
operation. The proposed changes are consistent with the safety
analysis assumptions and current plant operating practice. (Nuclear
Management Company, LLC) NMC is only requesting to revise the
existing numerical values and update the TS figure for the RCS P/T
limits based upon an NRC-approved methodology for calculating the
neutron fluence on the RPV, and to reflect a new limiting component.
The curves continue to be based upon ASME Code Case N-640, which has
been previously approved for use at the [Duane Arnold Energy Center]
DAEC.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
(3) The proposed amendment will not involve a significant
reduction in a margin of safety.
The proposed changes do not alter the manner in which Safety
Limits, Limiting Safety System Settings or Limiting Conditions for
Operation are determined. The setpoints at which protective actions
are initiated are not altered by the proposed changes. Sufficient
equipment remains available to actuate upon demand for the purpose
of mitigating an analyzed event. NMC is only requesting to revise
the existing numerical values and update the TS figure for the RCS
P/T limits based upon an NRC-approved methodology for calculating
the neutron fluence, NEDC-32983P-A. The new curves also reflect the
addition of a new limiting component, the recirculation inlet nozzle
(N2). No other changes to the Limiting Conditions for Operation or
any Surveillance Requirements of Technical Specification 3.4.9 are
proposed.
10 CFR 50, Appendix G specifies fracture toughness requirements
to provide adequate margins of safety during operation over the
service lifetime. The values of adjusted reference temperature and
upper shelf energy are expected to remain within the limits of
Regulatory Guide 1.99, Revision 2 and Appendix G of 10 CFR 50 for at
least 32 effective full power years (EFPY) of operation. The safety
analysis supporting this change continues to satisfy the ASME Code,
including ASME Code Case N-640, and 10 CFR 50, Appendices G and H
requirements and associated guidance documents, such as Regulatory
Guide 1.99, Rev. 2. Thus, the
[[Page 28856]]
proposed changes will not significantly reduce any margin of safety
that currently exists.
Based upon the above, NMC has determined that the proposed
amendment will not involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based upon
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: Jonathan Rogoff, Esquire, General Counsel,
NMC, LLC, 700 First St., Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: March 13, 2003.
Description of amendment request: The proposed amendment deletes
requirements from the Technical Specifications (TSs) and other elements
of the licensing bases to maintain a Post Accident Sampling System
(PASS). Licensees were generally required to implement PASS upgrades as
described in NUREG-0737, ``Clarification of TMI [Three Mile Island]
Action Plan Requirements,'' and Regulatory Guide 1.97,
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess
Plant and Environs Conditions During and Following an Accident.''
Implementation of these upgrades was an outcome of the lessons learned
from the accident that occurred at TMI, Unit 2. Requirements related to
PASS were imposed by Order for many facilities and were added to, or
included in, the TSs for nuclear power reactors currently licensed to
operate. Lessons learned and improvements implemented over the last 20
years have shown that the information obtained from PASS can be readily
obtained through other means, or is of little use in the assessment and
mitigation of accident conditions.
The changes are based on NRC-approved Technical Specification Task
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
413, ``Elimination of Requirements for a Post Accident Sampling System
(PASS).'' The NRC staff issued a notice of opportunity for comment in
the Federal Register on December 27, 2001 (66 FR 66949), on possible
amendments concerning TSTF-413, including a model Safety Evaluation and
model no significant hazards consideration (NSHC) determination, using
the consolidated line item improvement process. The NRC staff
subsequently issued a notice of availability of the models for
referencing in license amendment applications in the Federal Register
on March 20, 2002 (67 FR 13027). The licensee affirmed the
applicability of the following NSHC determination in its application
dated March 13, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
1. The proposed change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The PASS was originally designed to perform many sampling and
analysis functions. These functions were designed and intended to be
used in post-accident situations and were put into place as a result of
the TMI-2 accident. The specific intent of the PASS was to provide a
system that has the capability to obtain and analyze samples of plant
fluids containing potentially high levels of radioactivity, without
exceeding plant personnel radiation exposure limits. Analytical results
of these samples would be used largely for verification purposes in
aiding the plant staff in assessing the extent of core damage and
subsequent offsite radiological dose projections. The system was not
intended to, and does not, serve a function for preventing accidents
and its elimination would not affect the probability of accidents
previously evaluated.
In the 20 years since the TMI-2 accident, and the consequential
promulgation of post-accident sampling requirements, operating
experience has demonstrated that a PASS provides little actual benefit
to post-accident mitigation. Past experience has indicated that there
exists in-plant instrumentation and methodologies available in lieu of
a PASS for collecting and assimilating information needed to assess
core damage following an accident. Furthermore, the implementation of
Severe Accident Management Guidance (SAMG) emphasizes accident
management strategies based on in-plant instruments. These strategies
provide guidance to the plant staff for mitigation and recovery from a
severe accident. Based on current severe accident management strategies
and guidelines, it is determined that the PASS provides little benefit
to the plant staff in coping with an accident.
The regulatory requirements for the PASS can be eliminated without
degrading the plant emergency response. The emergency response, in this
sense, refers to the methodologies used in ascertaining the condition
of the reactor core, mitigating the consequences of an accident,
assessing and projecting offsite releases of radioactivity, and
establishing protective action recommendations to be communicated to
offsite authorities. The elimination of the PASS will not prevent an
accident management strategy that meets the initial intent of the post-
TMI-2 accident guidance through the use of the SAMGs, the emergency
plan (EP), the emergency operating procedures (EOP), and site survey
monitoring that support modification of emergency plan protective
action recommendations (PARs).
Therefore, the elimination of the PASS requirements from TSs (and
other elements of the licensing bases) does not involve a significant
increase in the consequences of any accident previously evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
The elimination of PASS-related requirements will not result in any
failure mode not previously analyzed. The PASS was intended to allow
for verification of the extent of reactor core damage, and also to
provide an input to offsite dose projection calculations. The PASS is
not considered an accident precursor, nor does its existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post-accident confinement of radioisotopes within the
containment building.
Therefore, this change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. The proposed change does not involve a significant reduction in
the margin of safety.
The elimination of the PASS, in light of existing plant equipment,
instrumentation, procedures, and programs that provide effective
mitigation of, and recovery from, reactor accidents, results in a
neutral impact to the margin of safety. Methodologies that are not
reliant on PASS are designed to provide rapid assessment of current
reactor core conditions and the direction of degradation while
effectively responding to the event in order to mitigate the
consequences of the accident. The use of a PASS is redundant and does
not provide quick recognition of core events or rapid response to
events in progress. The intent of the requirements established as
[[Page 28857]]
a result of the TMI-2 accident can be adequately met without reliance
on a PASS.
Therefore, this change does not involve a significant reduction in
the margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
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 21, 2003.
Description of amendment request: The proposed amendments would
revise Technical Specifications (TS) Section 5.5.1, ``Offsite Dose
Calculation Manual (ODCM),'' to remove reference to the Plant
Operations Review Committee review and acceptance of licensee initiated
changes to the ODCM.
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?
The proposed change for TS section 5.5.1.b removes the reference
to the Plant Operations Review Committee review and acceptance of
licensee initiated changes to the ODCM. This change is an
administrative change and does not change plant design or responses.
The proposed change does not involve changing any structure,
system, or component, or affect reactor operations. It is not an
initiator of an accident and does not change any existing safety
analysis previously analyzed in the UFSAR. As such, the proposed
change does not involve a significant increase in the probability of
an accident previously evaluated. Since the proposed change does not
alter the plant design, it 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?
The proposed change for TS section 5.5.1.b removes the reference
to the Plant Operations Review Committee review and acceptance of
licensee initiated changes to the ODCM. This change is an
administrative change and does not change plant design or responses.
The proposed change will not alter any plant design basis or
postulated accident. In addition, the proposed change does not
impact any plant systems or components.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed change for TS section 5.5.1.b removes the reference
to the Plant Operations Review Committee review and acceptance of
licensee initiated changes to the ODCM. This change is an
administrative change and does not change plant design or responses.
The proposed change does not impact accident offsite dose,
containment pressure or temperature, emergency core cooling system
setpoints, reactor protection system settings or any other parameter
that could affect 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: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201.
NRC Section Chief: John A. Nakoski.
Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260 and 50-
296, Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,
Alabama
Date of amendment request: April 11, 2003 (TS-424).
Description of amendment request: The proposed amendments would
reduce the number of Emergency Core Cooling System subsystems that are
available in response to certain design basis loss-of-coolant accident
(LOCA) scenarios because of TVA's planned restart of Unit 1. The
licensee stated that the reduced number has been analyzed and is
consistent with the current approved LOCA analysis methodology. The
amendments are needed to eliminate the potential for overloading a
shutdown board or a diesel generator when both Units 1 and 2 are in-
service. The reduction requires a change to the Updated Final Safety
Analysis Report.
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 amendments and Technical Specification
changes involve a significant increase in the probability or
consequences of an accident previously evaluated?
No. The proposed amendments revise the actual number of
Emergency Core Cooling System (ECCS) subsystems that are available
in response to certain design basis Loss of Coolant Accident (LOCA)
scenarios. The associated modifications also result in a revision to
the number of required channels for the Low Pressure Coolant
Injection (LPCI) pump start time delay relay function specified in
Technical Specifications. The proposed amendments and Technical
Specification changes do not affect any accident precursors.
Therefore, the probability of an evaluated accident is not
increased.
The reduction in the number of ECCS subsystems that are actually
available in response to the bounding LOCA case (A recirculation
suction line break with an assumed battery failure) will now be the
same as the number of ECCS subsystems evaluated in the current BFN
SAFER/GESTR-LOCA analysis. The ECCS performance for the bounding
LOCA case has previously been evaluated using the approved SAFER/
GESTR-LOCA application methodology and is described in Updated Final
Safety Analysis Report (UFSAR) Sections 6.5 and 14.6.3. The revision
to the number of required channels for the LPCI pump start time
delay relay function does not affect the LOCA analysis. The
requirements of 10 CFR 50.46 and Appendix K are met. Therefore, the
proposed amendments and Technical Specification changes will not
significantly increase the consequences of an accident previously
evaluated.
2. Do the proposed amendments and Technical Specification
changes create the possibility of a new or different kind of
accident from any accident previously evaluated?
No. The proposed amendments revise the number of ECCS subsystems
that are actually available in response to certain design basis LOCA
scenarios. The proposed Technical Specification changes revise the
number of required channels for the LPCI pump start time delay relay
function. The proposed amendments and Technical Specification
changes do not introduce new equipment, which could create a new or
different kind of accident.
No new external threats, release pathways, or equipment failure
modes are created. Therefore, the implementation of the proposed
amendments and Technical Specification changes will not create a
possibility for an accident of a new or different type than those
previously evaluated.
3. Do the proposed amendments and Technical Specification
changes involve a significant reduction in a margin of safety?
No. The proposed amendments and Technical Specification changes
revise the number of ECCS subsystems that are actually available in
response to certain design basis LOCA scenarios. The reduction in
the number of ECCS subsystems that are actually available in
response to the bounding LOCA case (A recirculation suction line
break with an assumed battery failure) will now be the same as the
number of ECCS subsystems evaluated in the current BFN SAFER/GESTR-
LOCA analysis. The ECCS performance for the bounding LOCA case has
previously been evaluated using the approved SAFER/GESTR-LOCA
application methodology. The revision to the number of required
channels for the LPCI pump start time delay relay function does not
affect the LOCA analysis. The requirements of 10 CFR 50.46 and
Appendix K are met. Therefore,
[[Page 28858]]
the proposed license amendments and Technical Specification 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Allen G. Howe.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
Browns Ferry Nuclear Plant (BFN), Units 1, 2 and 3, Limestone County,
Alabama
Date of amendment request: April 15, 2003 (TS 409).
Description of amendment request: The proposed amendments are
applicable to BFN Units 1, 2, and 3. They would revise Technical
Specification (TS) Limiting Condition for Operation 3.7.3, Control Room
Emergency Ventilation (CREV) System, and its associated TS Bases to
provide specific conditions and required actions that address a
degraded main control room boundary. The proposed changes are
consistent with the TS Task Force Traveler 287, Revision 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. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
2. No. The proposed TS change involves the CREV system, which
provides a radiological controlled environment from which the plant
can be operated following a design basis accident (DBA). The CREV
system is not assumed to be the initiator of any analyzed accident
and cannot not [sic] affect the probability of accidents.
The proposed change allows the main control room boundary to be
opened intermittently under administrative control, and allows 24
hours to restore the main control room boundary to Operable status
before requiring the plant to perform an orderly shutdown. The 24-
hour Completion Time is reasonable based on the low probability of a
DBA occurring during this time period and TVA's commitment to
implement, via administrative controls, appropriate compensatory
measures consistent with the intent of 10 CFR part 50, Appendix A,
General Design Criteria (GDC) 19. These compensatory measures
minimize the consequences of an open main control room boundary and
assure that CREV system can continue to perform its function. As
such, these changes will not affect the function or operation of any
other systems, structures, or components.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
No. The proposed change allows the main control room boundary to
be opened intermittently under administrative control and allows 24
hours to restore the main control room boundary to Operable status
before requiring the plant to perform an orderly shutdown. The 24-
hour Completion Time is reasonable based on the low probability of a
DBA occurring during this time period and TVA's commitment to
implement, via administrative controls, appropriate compensatory
measures consistent with the intent of 10 CFR part 50, Appendix A,
GDC 19. These compensatory measures minimize the consequences of an
open main control room boundary and assure that the CREV system can
continue to perform its function. As such, these changes will not
affect the function or operation of any other systems, structures,
or components.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No. The proposed change allows the main control room boundary to
be opened intermittently under administrative control and allows 24
hours to restore the main control room boundary to Operable status
before requiring the plant to perform an orderly shutdown. The 24-
hour Completion Time is reasonable based on the low probability of a
DBA occurring during this time period and TVA's commitment to
implement, via administrative controls, appropriate compensatory
measures consistent with the intent of 10 CFR part 50, Appendix A,
GDC 19. These compensatory measures minimize the consequences of an
open main control room boundary and assure that the CREV system can
continue to perform its function such that compliance with GDC 19 is
maintained.
Therefore, the proposed TS change does not involve a reduction
in the margin of safety.
Based on the above, TVA 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
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: Allen G. Howe.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant (BFN), Units 2 and 3, Limestone County, Alabama
Date of amendment request: April 14, 2003 (TS 425).
Description of amendment request: The proposed amendments would
revise two Technical Specification (TS) Limiting Conditions for
Operation 3.3.4.1, ``End Of Cycle Recirculation Pump Trip (EOC-RPT)
Instrumentation,'' and 3.7.5, ``Main Turbine Bypass System,'' to
reference additional core limits adjustment factors for linear heat
generation rate for equipment out-of-service conditions. Also, Section
b of TS 5.6.5, ``Core Operating Limits Report (COLR),'' would be
revised to add references to the Framatome Advanced Nuclear Power
(FANP) analytical methods that will be used in the upcoming fuel cycles
to determine core operating limits. The above TS changes are needed to
support a transition to the use of FANP fuel, and FANP core design and
analysis services.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. Core operating limits are established to support
requirements, which in turn ensure that fuel design limits are not
exceeded during any conditions of operating transients or accidents.
The methods used to determine the limits for each operating cycle
are based on methods previously found acceptable by the NRC and are
required to be listed in COLR TS Section 5.6.5.b. Accordingly, a
change to TS Section 5.6.5.b is requested to include FANP methods in
the list of NRC-approved methods applicable to BFN. This TS change
also adds provisions that ensure core thermal limits adjustment
factors are applied for equipment out-of-service conditions
associated with the use of FANP methods for transient analyses. The
application of these NRC-approved methods will continue to ensure
that acceptable operating limits are established and applied for
protection of fuel cladding integrity during transient and
accidents.
The requested TS changes do not involve any plant modifications
or operational changes that could affect system reliability,
[[Page 28859]]
performance, or possibility of operator error. The requested changes
do not affect any postulated accident precursors, do not affect any
accident mitigation systems, and do not introduce any new accident
initiation mechanisms.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
No. The core operating limits and required limits adjustments
for equipment out-of-service conditions will continue to be
determined using methodologies that have been approved by the NRC.
The limits derived from approved methodologies will provide adequate
margins of safety. The proposed changes do not involve any new modes
of operation, any changes to setpoints, or any plant modifications,
and do not result in any new precursors to an accident.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety.
No. The core operating limits and required limits adjustments
for equipment out-of-service will continue to be determined using
methodologies that have been approved by the NRC. On this basis, the
implementation of the changes does not involve a significant
reduction in 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: Allen G. Howe.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant
(WBN), Unit 1, Rhea County, Tennessee
Date of amendment request: May 1, 2003.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.8.7, ``Inverters--Operating.''
The TS as currently written requires two inverters for each of the four
instrument channels. The revision changes the requirement to one
inverter for each of the four channels. The amendment is the initial
phase of a project that will replace the vital inverters to achieve
improvements in the reliability of the 120V AC Vital Instrument Power
System.
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?
No. The proposed revisions to WBN's Vital AC Power System do not
alter the safety functions of the Vital Inverters or the Unit 1 and
Unit 2 120V AC Vital Instrument Power Boards. The initial conditions
for the Design Basis Accidents (DBAs) defined in the WBN Updated
Final Safety Analysis Report (UFSAR) assume the Engineered Safety
Feature (ESF) systems are operable. The vital inverters are designed
to provide the required capacity, capability, redundancy, and
reliability to ensure the availability of necessary power to vital
instrumentation so that the fuel, reactor coolant system, and
containment design limits are not exceeded. Adding the Unit 2 loads
to the Unit 1 inverters does not alter the accident analyses as long
as the Unit 1 inverters are capable of handling the additional loads
and channel separation is maintained. Design calculations document
that the Unit 1 inverters have adequate capacity to support the
addition of the Unit 2 loads and no changes are proposed that will
impact the separation of the Vital AC Power System. In addition, the
redundant capabilities of the Vital AC System as currently described
in the UFSAR are not impacted by the proposed amendment.
The inverters and the associated 120V AC Vital Instrument Power
Boards are utilized to support instrumentation that monitor critical
plant parameters to aid in the detection of accidents and to support
the mitigation of accidents, but are not considered to be an
initiator of design basis accidents. Based on this and the preceding
information, the proposed amendment 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?
No. When implemented, the proposed TS amendment will allow the
Unit 2 Vital Instrument Power Boards to receive their
uninterruptible power supply (UPS) power from the Unit 1 inverters
instead of the Unit 2 inverters. Calculations have verified that the
additional load will not affect the ability of the Unit 1 inverters
to perform their intended safety functions. In addition, the
inverters and the 120V AC Vital Instrument Power Boards are not
considered to be an initiator of a design basis accident. These
components provide power to instrumentation that supports the
identification and mitigation of accidents as well as system control
functions during normal plant operations. The functions of the
inverters are not altered by the proposed TS change and will not
create the possibility of a new or different accident. Further, the
addition of the Unit 2 loads to the Unit 1 inverters is the
principal change to the inverter system and this change is bounded
by previously evaluated accident analyses. Therefore, the proposed
amendment does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The plant setpoints and limits that are utilized to ensure
safe operation and detect accident conditions are not impacted by
the proposed TS amendment. The inverters and the 120V Vital
Instrument Power Boards will continue to provide reliable power to
safety-related instrumentation for the identification and mitigation
of accidents and to support plant operation. Therefore, the margin
of safety is not reduced.
Based on the above, TVA 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.
In conclusion, based on the considerations discussed above, (1)
There is reasonable assurance that the health and safety of the
public will not be endangered by operation in the proposed manner,
(2) such activities will be conducted in compliance with the
Commission's regulations, and (3) the issuance of the amendment will
not be inimical to the common defense and security or to the health
and safety of the public.
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: Allen G. Howe.
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
[[Page 28860]]
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 email to [email protected].
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: August 1, 2002, as supplemented
on October 18, 2002, and April 17, 2003.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.7.1.1, ``Plant Systems: Turbine Cycle Safety
Valves,'' to reflect results of a reanalysis of overpressurization
events to allow plant operation, at corresponding reduced power levels,
with up to four main steam safety valves in each main steam line
inoperable.
Date of issuance: May 7, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 275.
Facility Operating License No. DPR-65: This amendment revised the
TSs.
Date of initial notice in Federal Register: September 17, 2002 (67
FR 58638). The supplements dated October 18, 2002, and April 17, 2003,
provided additional information which clarified the application, did
not expand the scope of the application as originally noticed, and did
not change the staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 7, 2003.
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: September 30, 2002, as
supplemented by letters dated October 17, 2002 and April 2, 2003.
Brief description of amendments: The amendments revise the
Technical Specification to: (1) Modify the Surveillance Requirement to
be consistent with the design of the reactor building access openings,
(2) modify the frequency of the Surveillance Requirement for visual
inspections for the exposed interior and exterior surface of the
reactor building, and (3) modify the administrative controls for the
containment leakage rate testing program.
Date of issuance: May 8, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: 212/193.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 12, 2002 (67
FR 68733). The supplement dated October 17, 2002, and April 12, 2003,
provided clarifying information that did not change the scope of the
September 30, 2002, application nor the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 8, 2003.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: September 3, 2002, as
supplemented by letters dated November 27, 2002, and April 17, 2003.
Brief description of amendment: The amendment allows the addition
of depleted uranium to the fuel assembly composition described in
Technical Specification (TS) 4.2.1. The amendment also revises TS
5.6.5.b to incorporate the references to the analytical methods to be
used to determine core operating limits and removes those references
that will no longer be used. The amendment also allows the format for
those document references to be revised as described in the staff-
approved Industry/TSTF Standard Technical Specification Change
Traveler, TSTF-363, ``Revise Topical Report References in ITS 5.6.5,
COLR.''
Date of issuance: May 12, 2003.
Effective date: May 12, 2003, and shall be implemented within 30
days from the date of issuance.
Amendment No.: 185.
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 15, 2002 (67 FR
63693). The November 27, 2002, and April 17, 2003, supplemental letters
provided additional clarifying information, did not change the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 12, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: June 26, 2002, as supplemented
on March 12, 2003.
Brief description of amendment: The amendment revises Technical
Specification 5.6.5.b, ``Core Operating Limits Report (COLR),'' to
incorporate the reference to Westinghouse Topical Report WCAP-12945-P-
A, ``Code Qualification Document for Best Estimate Loss-of-Coolant
Analysis,'' dated March 1998. The amendment allows the use of the
analytical methodology to determine the core operating limits.
Date of issuance: May 6, 2003.
Effective date: May 6, 2003.
Amendment No.: 217.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 6, 2002 (68 FR
50952). The March 12 letter provided clarifying information that did
not expand the scope of the Federal Register notice or change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is
[[Page 28861]]
contained in a Safety Evaluation dated May 6, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: July 5, 2002, as supplemented on
September 27, November 6, November 21, and December 30, 2002; February
4, February 10, March 17, and April 14, 2003.
Brief description of amendment: The amendment increases the
licensed power level by 1.5%, from 1998 MWt to 2028 MWt, based on the
installation of ultrasonic flow measurement instrumentation resulting
in improved feedwater flow measurement accuracy. The amendment changes
the Operating License (OL) and Technical Specifications (TSs) to
reflect the increase in licensed power level.
Date of issuance: May 9, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 201.
Facility Operating License No. DPR-35: Amendment revised the TSs
and OL.
Date of initial notice in Federal Register: September 3, 2002 (67
FR 56322). The supplements dated September 27, November 6, November 21,
and December 30, 2002; February 4, February 10, March 17, and April 14,
2003, provided additional information that clarified the application,
and did not expand the scope of the application or change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated May 9, 2003.
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: February 27, 2003, as
supplemented April 7, 2003.
Brief description of amendments: The amendments revise the
Technical Specifications by adding a surveillance requirement to
perform a quarterly trip unit calibration of the reactor protection
system scram discharge volume water level--high differential pressure
switches.
Date of issuance: May 6, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 214/208.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 1, 2003 (68 FR
15760). The supplement dated April 7, 2003, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated May 6, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket Nos. 50-334 and
50-412, Beaver Valley Power Station, Unit Nos. 1 and 2, Beaver County,
Pennsylvania
Date of application for amendments: January 16, 2002, as
supplemented October 17, 2002.
Brief description of amendments: These amendments revised portions
of the current Technical Specifications, Section 6.0, ``Administrative
Controls,'' to conform with improved Technical Specifications. The
conversion is based upon: NUREG-1431, ``Standard Technical
Specifications for Westinghouse Plants,'' Revision 2, dated April 2001;
``Final Policy Statement on Technical Specification Improvements for
Nuclear Power Reactors'' (Final Policy Statement), published on July
22, 1993 (58 FR 39132); and Title 10 of the Code of Federal Regulations
(10 CFR), Sec. 50.36, ``Technical Specifications,'' as amended July
19, 1995.
Date of issuance: May 15, 2003.
Effective date: As of date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 255 and 136.
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 29, 2002 (67 FR
66010). The supplement dated October 17, 2002, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the Nuclear
Regulatory Commission (NRC) staff's original proposed 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 15, 2003.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-412,
Beaver Valley Power Station, Unit 2, Beaver County, Pennsylvania
Date of application for amendment: May 31, 2002, as supplemented
September 11, 2002, January 30, and February 21, 2003.
Brief description of amendment: The amendment revised the Technical
Specification Design Feature 5.3.1, Criticality, such that the new fuel
(fresh fuel) racks enrichment limit specified in Section 5.3.1.2.a was
increased from 4.85 weight percent to a 5.00 weight percent limit.
Date of issuance: May 15, 2003.
Effective date: As of date of issuance and shall be implemented
within 60 days.
Amendment No: 135.
Facility Operating License No. NPF-73. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 17, 2002 (67
FR 58645). The September 11, 2002, January 30, and February 21, 2003,
letters provided additional information that clarified the application,
did not expand the scope of the application as originally noticed, and
did not change the original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 15, 2003.
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: March 22, 2002, as supplemented May 13,
June 24, July 29, and December 20, 2002.
Description of amendment request: The amendment revises Technical
Specifications (TSs) Surveillance Requirement (SR) 4.0.3 to extend the
delay period, before entering a Limiting Condition for Operation,
following a missed surveillance. The delay period is extended from the
current limit of ``... up to 24 hours'' to ``...up to 24 hours or up to
the limit of the specified surveillance interval, whichever is
greater.'' In addition, the following requirement is added to SR 4.0.3:
``A risk evaluation shall be performed for any Surveillance delayed
greater than 24 hours and the risk impact shall be managed.'' The
amendment also adds a requirement for a TS Bases Control Program to the
administrative controls section of TSs and makes administrative changes
to SRs 4.0.1 and 4.0.3 to be consistent with NUREG-1431, Revision
[[Page 28862]]
2, ``Standard Technical Specifications Westinghouse Plants.''
Date of issuance: May 15, 2003.
Effective date: As of its date of issuance, and shall be
implemented within 30 days.
Amendment No.: 87.
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 21, 2003 (68 FR
2804).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 15, 2003.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook
Nuclear Plant, Unit 2, Berrien County, Michigan
Date of application for amendment: November 15, 2002, as
supplemented February 24 and April 25, 2003.
Brief description of amendment: The amendment increases the
licensed reactor core power level by 1.66 percent from 3411 megawatts
thermal (MWt) to 3468 MWt. The power level increase is considered a
measurement uncertainty recapture power uprate.
Date of issuance: May 2, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 259.
Facility Operating License No. DPR-74: Amendment revises the
Operating License and Technical Specifications.
Date of initial notice in Federal Register: January 21, 2003 (68 FR
2805)
The February 24 and April 25, 2003, supplemental letters provided
additional clarifying information that was within the scope of the
original application and did not change the Nuclear Regulatory
Commission staff's initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 2, 2003.
No significant hazards consideration comments received: No.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: January 27, 2003.
Brief description of amendment: The amendment makes administrative
and editorial changes to the Fort Calhoun Station Technical
Specifications 1.3 Basis (1); 2.7 (1)a; 2.7 (1)b; 2.7 (1)d; 2.7 (1)i;
2.7 Basis; 3.0.2; Table 3-5, Item 11; and 3.5(3)ii. The changes are
primarily editorial and are typographical changes or corrections.
Date of issuance: May 8, 2003.
Effective date: May 8, 2003, and shall be implemented within 60
days from the date of issuance.
Amendment No.: 218.
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 18, 2003 (68 FR
12955).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 8, 2003.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket No. 50-206, San
Onofre Nuclear Generating Station, Unit 1, San Diego County, California
Date of application for amendment: March 11, 2003.
Brief description of amendments: The amendment application requests
a revision to the Unit 1 defueled Technical Specifications
administrative controls section to propose changes in organizational
responsibilities. Specifically, the proposed changes identify that the
Vice President, Engineering & Technical Services will be responsible
for decommissioning activities. Additionally, the Station Manager will
be designated as having approval authority for activities within the
Station Manager's organization.
Date of issuance: May 15, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: Unit 1-161.
Facility Operating License No. DPR-13: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18285).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 15, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: February 19, 2003.
Description of amendment request: The amendments deleted Technical
Specification 5.5.3, ``Post Accident Sampling'' and, thereby,
eliminated the requirements to have and maintain the post accident
sampling system.
Date of issuance: May 9, 2003.
Effective date: Date of issuance, to be implemented within 60 days.
Amendment Nos.: 245, 282, 240.
Facility Operating License Nos. DPR-33, DPR-52, and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 18, 2003 (68 FR
12957).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated May 9, 2003.
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: July 25, 2002, as supplemented by
letters dated February 5 and February 11, 2003.
Brief description of amendments: The amendments change the Comanche
Peak Steam Electric Station, Units 1 and 2, Facility Operating Licenses
as follows: The license conditions related to Decommissioning Trusts,
specified in Sections 2.C.(4)(a), 2.C.(4)(b), 2.C.(4)(d), 2.C(4)(e),
and 2.C.(6), are deleted and Section 2.E, which requires reporting any
violations of the requirements contained in Section 2.C of the
licenses, is deleted. Additionally, Technical Specification Table 5.5-
2, ``Steam Generator Tube Inspection,'' Table 5.5-3, ``Steam Generator
Repaired Tube Inspection for Unit 1 Only,'' and TS 5.6.10c, ``Steam
Generator Tube Inspection Report,'' are revised to delete the
requirement to notify the NRC pursuant to Sec. 50.72(b)(2),
``Immediate notification requirements for operating nuclear power
reactors,'' of Title 10 of the Code of Federal Regulations (10 CFR) if
the steam generator tube inspection results are in a Category C-3
classification.
Date of issuance: May 15, 2003.
Effective date: December 24, 2003, and shall be implemented within
60 days from that date.
Amendment Nos.: 103/103.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 3, 2002 (67
FR 56329).
The February 5, 2003, supplement was the subject of a second no
significant hazards consideration determination (68 FR 10282, published
March 4, 2003). The February 11, 2003, supplement provided clarifying
information that did not change the scope of the original Federal
Register notice or the original no significant hazards consideration
determination.
[[Page 28863]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 15, 2003.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, 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 Public
Document Room (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. By June 26, 2003, 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 2.714, 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 an Atomic Safety and Licensing Board, designated by the
Commission or by the Chairman of the Atomic Safety and Licensing Board
Panel, will rule on the request and/or petition; and the Secretary or
the designated Atomic Safety and Licensing Board will issue a notice of
a hearing or an appropriate order.
As required by 10 CFR 2.714, 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 factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the
[[Page 28864]]
proceeding, but such an amended petition must satisfy the specificity
requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of 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. Petitioner must provide 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 to relief. A
petitioner who fails to file such a supplement which satisfies 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, including the opportunity to present evidence and cross-
examine witnesses. 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 with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of the continuing disruptions in delivery of mail to United
States Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the petition for
leave to intervene and request for hearing should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, 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 filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
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: April 25, 2003.
Brief description of amendments: The amendments modify Technical
Specification surveillance requirements to provide an alternative means
of testing the Unit 2 main steam power operated relief valves,
including those that provide the automatic depressurization system and
low set relief functions.
Date of issuance: May 8, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 215/209.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes. Quad-City Times, dated May 5, 2003. The
notice provided an opportunity to submit comments on the Commission's
proposed NSHC determination. No comments have been received.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, state consultation, and final NSHC determination
are contained in a Safety Evaluation dated May 8, 2003.
Dated at Rockville, Maryland, this 19th day of May 2003.
For the Nuclear Regulatory Commission.
William H. Ruland,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 03-12973 Filed 5-23-03; 8:45 am]
BILLING CODE 7590-01-U