[Federal Register Volume 66, Number 220 (Wednesday, November 14, 2001)]
[Notices]
[Pages 57116-57131]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-28399]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
Note: The publication date for this notice will change from
every other Wednesday to every other Tuesday, effective January 8,
2002. The notice will contain the same information and will continue
to be published biweekly.
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
[[Page 57117]]
make immediately effective any amendment to an operating license upon a
determination by the Commission that such amendment involves no
significant hazards consideration, notwithstanding the pendency before
the Commission of a request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from October 22, 2001 through November 3, 2001.
The last biweekly notice was published on October 31, 2001 (66 FR
557007).
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 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 NRC Public Document
Room, located at One White Flint North, 11555 Rockville Pike (first
floor), Rockville, Maryland. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By December 14, 2001, 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 NRC's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland 20852.
Publicly available records will be accessible electronically 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/NRC/ADAMS/index.html. 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.
[[Page 57118]]
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 Branch, or may be delivered to the Commission's Public
Document Room, located at One White Flint North, 11555 Rockville Pike
(first floor), Rockville, Maryland 20852, by the above date. A copy of
the petition should also be sent to the Office of the General Counsel,
U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, and 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 public inspection at
the Commission's Public Document Room, located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Assess and Management Systems (ADAMS) Public Electronic
Reading Room on the internet at the NRC web site, http://www.nrc.gov/NRC/ADAMS/index.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, 304-
415-4737 or by email to [email protected].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit 1, DeWitt County, Illinois
Date of amendment request: August 21, 2001.
Description of amendment request: The proposed amendment would
revise the actions taken for an inoperable battery charger, revise
battery charger testing criteria, and relocate certain safety-related
battery surveillance requirements from the Technical Specifications to
a licensee-controlled 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:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes restructure the TS [Technical
Specifications] for the DC Electrical Power system. The proposed
changes add actions to specifically address battery charger
inoperability with increased completion times. This change will rely
upon the capability of providing the battery charger function by an
alternate means, (e.g., a spare battery changer that will function
as a qualified backup) to take advantage of the proposed increased
completion time. The CD power System or associated battery chargers
are not initiators to any accident sequence analyzed in the Updated
Safety Analysis Report (USAR). Operation in accordance with the
proposed TS ensures that the DC Power System is capable of
performing function as described in the USAR, therefore the
mitigative functions supported by the DC Power System will continue
to provide the protection assumed by the analysis.
The relocation of preventive maintenance surveillance, and
certain operating limits and actions to a newly-created, licensee-
controlled TS 5.5.14, ``Battery Monitoring and Maintenance
Program,'' will not challenge the ability of the DC Power System to
perform its design function. The maintenance and monitoring required
by current TS, which are based on industry standards, will continue
to be performed. In addition, the DC Power System is within the
scope of 10 CFR 50.65, ``Requirements for monitoring the
effectiveness of maintenance at nuclear power plants,'' which will
ensure the control of maintenance activities associated with the DC
Power System.
In summary, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes involve restructuring the TS for the DC
Electrical Power system. This change will rely upon the capability
of providing the battery charger function by an alternate means,
(e.g., a spare battery charger that will function as a qualified
backup) to take advantage of the proposed increased completion time.
The DC Power System or associated battery chargers are not
initiators to any accident sequence analyzed in the Updated Safety
Analysis Report (USAR).
Allowing the use of a spare battery charger will increase the
reliability of the DC Electrical Power system. The mitigative
functions supported by the DC Power System will continue to provide
the protection assumed by the safety analysis described in the USAR.
Therefore, there are no new types of failures that could be created
by a failure of the spare battery charger. As such, no new or
different kind of accident or transient is expected by these
changes.
Therefore, these proposed changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Does the change involve a significant reduction in a margin of
safety?
The proposed changes will not adversely affect operation of
plant equipment. These changes will not result in a change to the
setpoints at which protective actions are initiated. Sufficient DC
capacity to support operation of mitigation equipment is ensured.
The changes associated with the new Battery Maintenance and
Monitoring Program will ensure that the station batteries are
maintained in a highly reliable manner. The use of a spare battery
charger will increased the reliability of the DC system during
periods of normal battery charger inoperability. The equipment fed
by the DC Electrical Sources will continue to provide adequate power
to safety related loads in accordance with analysis assumptions.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Robert Helfrich, Mid-West Regional Operating
Group, Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville,
IL 60555.
NRC Section Chief: Anthony J. Mendiola.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: September 20, 2000, as supplemented
August 2 and September 28, 2001.
Description of amendment request: The proposed Technical
Specification (TS) change would (1) delete the requirements for
hydrogen monitoring instrumentation from TS sections 3.5.5.2, 3.6, and
Tables 3.5-3 and 4.1-4 and correct a typographical error in item 8 of
Table 4.1-4; (2) delete the requirements for hydrogen recombiners in TS
section 4.4.4; and (3) delete the reference to the hydrogen purge
system and hydrogen recombiners from the Bases of TS section 4.12.2.
Basis for proposed no significanthazards consideration
determination: As required by 10 CFR
[[Page 57119]]
50.91(a), the licensee has provided its analysis of the issue of no
significant hazards consideration. The Nuclear Regulatory Commission
(NRC) staff reviewed the licensee's analysis against the standards of
10 CFR 50.92(c). The NRC staff's analysis, which is based on the
representation made by the licensee in the September 20, 2001,
application as supplemented August 2 and September 28, 2001, is
presented below:
1. Will operation of the facility in accordance with this proposed
change involve a significant increase in the probability or
consequences of an accident previously evaluated?
No. This change has no effect on plant equipment provided for the
reactor coolant system, reactor building heat removal, or the equipment
provided for mixing of the reactor building atmosphere following an
accident. This proposed change does not alter the design or
configuration of the plant beyond that of the containment combustible
gas control systems. The containment combustible gas control systems
are currently classified as safety systems. The containment combustible
gas control systems are composed of two hydrogen monitors and two
hydrogen recombiners, backed up by a portion of the reactor building
purge system that can be used to vent the reactor building. Hydrogen
control components (hydrogen monitors, hydrogen recombiners, and
hydrogen vents) do not affect any accident initiation sequence
previously identified. Therefore, this change does not increase the
probability of an accident previously evaluated.
The containment combustible gas control systems are provided to
ensure that reactor building hydrogen concentration is maintained below
the lower flammability limit of 4.0 percent. The NRC staff has found
hydrogen combustion to be a small contributor to containment failure
for large, dry containment designs due to the robustness of these
containment types and the likelihood of a spurious ignition source. The
containment combustible gas control systems are not credited in the TMI
Unit 1 probability risk assessment (PRA).
Therefore, this change would not result in a significant increase
the consequence of accidents previously evaluated.
2. Will operation of the facility in accordance with the proposed
change create the possibility of a new or different kind of accident
from any accident previously evaluated?
No. This proposed change does not alter the design or configuration
of the plant beyond that of the containment combustible gas control
systems. Hydrogen generation following a design basis loss-of-coolant
accident (LOCA) has been evaluated in accordance with regulatory
requirements. Deletion of the containment combustible gas control
system from the TSs does not alter the hydrogen generation processes
post-LOCA. The NRC staff has found hydrogen combustion to be a small
contributor to containment failure for large, dry containment designs
due to the robustness of these containment types and the likelihood of
a spurious ignition source. The containment combustible gas control
systems are not credited in the TMI Unit 1 level 2 PRA.
Therefore, since the accident evaluation does not credit these
systems or assume that they operate during an accident, operation of
the facility in accordance with this proposed change will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Will operation of the facility in accordance with this proposed
change involve a significant reduction in a margin of safety?
No. This change has no effect on plant equipment provided for the
reactor coolant system, reactor building heat removal, or the equipment
provided for mixing of the reactor building atmosphere following an
accident. This change only involves the deletion of requirements for
containment combustible gas control equipment, (hydrogen monitors,
hydrogen recombiners, and containment hydrogen vents). The NRC staff
has found hydrogen combustion to be a small contributor to containment
failure for large, dry containment designs due to the robustness of
these containment types and the likelihood of a spurious ignition
source. Use of the containment combustible gas control systems are not
credited in the TMI Unit 1 PRA. TMI Unit 1 utilizes a large open
containment design that precludes the buildup of hydrogen pockets that
might be formed if the reactor building were of a compartmentalized
design. The TMI-1 PRA concluded that the containment would remain
intact for severe accidents which included hydrogen burns for which no
credit was taken for the combustible gas control system as long as the
containment heat removal systems (reactor building emergency cooling
and reactor building sprays) remain functional.
The proposed change will relax certain special treatment
requirements associated with hydrogen monitors. As discussed in
Regulatory Guide 1.97, ``Instrumentation for Light-Water-Cooled Nuclear
Power Plants to Assess Plant and Environs Conditions During and
Following an Accident,'' Revision 3, dated May 1983, the NRC staff
believes that the revised treatment is appropriate for instrumentation
needed to assess the degree of core damage and confirm that spurious
ignition of hydrogen has taken place.
Therefore, operation of the facility in accordance with this
proposed change will 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: Edward J. Cullen, Jr., Vice President and
General Counsel, Exelon Generation Company, LLC, 300 Exelon Way, KSB 3-
W, Kennett Square, PA 19348.
NRC Section Chief: L. Raghavan, Acting.
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: August 27, 2001.
Description of amendment request: The proposed amendment changes
the Millstone Nuclear Power Station, Unit No. 3 (MP3) Technical
Specifications (TSs) action and surveillance requirements associated
with the containment air lock. The Bases of the affected TSs will be
modified to address the proposed changes.
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 reviewed the licensee's analysis against
the standards of 10 CFR 50.92(c). The NRC staff's analysis, which is
based on the representation made by the licensee in the August 27,
2001, application, is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes will not revise the operability requirements
for the containment air lock. As a result, the design-basis accidents
will remain the same postulated events, and the consequences of the
design-basis accidents will remain the same. Also, the containment air
lock is not an accident initiator. Therefore, the proposed change will
not involve any increase in the probability or consequences of an
accident previously evaluated.
[[Page 57120]]
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
Since the containment air lock is not an accident initiator, these
proposed changes do not introduce any new failure modes. Therefore, the
proposed changes will not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
Since the operability requirements for the containment air lock
will not change, and the containment air lock will continue to function
as assumed in the safety analysis, the proposed change will not result
in a reduction in a margin of safety.
Based on this 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: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
NRC Section Chief: James W. Clifford.
Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No 3, New London County, Connecticut
Date of amendment request: September 26, 2001.
Description of amendment request: The proposed amendment modifies
the Millstone Nuclear Power Station, Unit No. 3 (MP3) Technical
Specifications (TSs) to relocate MP3 TSs related to the position
indication system to the respective Technical Requirements Manual
(TRM). The Bases of the affected TSs will be modified to address the
proposed changes. Also, index pages will be revised to reflect the
relocation.
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 reviewed the licensee's analysis against
the standards of 10 CFR 50.92(c). The NRC staff's analysis, which is
based on the representation made by the licensee in the September 26,
2001, application, is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed requirements remain the same except that the
requirements will be relocated to the TRM. Since the proposed
requirements are the same, this proposed 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.
Since the requirements remain the same, these proposed changes do
not alter the way any system, structure, or component functions and do
not alter the manner in which the plant is operated. The proposed
changes do not introduce any new failure modes. Therefore, the proposed
changes will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
Since the proposed changes are solely to relocate the existing
requirements, it does not affect plant operation in any way. Therefore,
the proposed change will not result in a reduction in a margin of
safety.
Based on this 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: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Waterford, CT 06141-5127.
NRC Section Chief: James W. Clifford.
Dominion Nuclear Connecticut Inc., et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of amendment request: October 1, 2001.
Description of amendment request: The proposed amendment modifies
the Millstone Nuclear Power Station, Unit No. 3 (MP3) Technical
Specifications (TS) to change TS 3.4.6.2 ``Reactor Coolant System--
Operational Leakage''. The Bases for this TS will also be modified to
reflect this change.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to Technical Specification 3.4.6.2 for
[reactor coolant systems] RCS PIVs [pressure isolation valve] in the
RHR [residual heat removal] flow path will not cause an accident to
occur and will not result in any change in the operation of
associated accident mitigation equipment. The ability of the RHR
System to remove core decay heat will not be affected. The proposed
changes will not affect the ability of the RCS or the RHR System to
mitigate any design basis event. The design basis accidents will
remain the same postulated events described in the Millstone Unit
No. 3 Final Safety Analysis Report (FSAR), and the consequences of
the design basis accidents will remain the same. Therefore, the
proposed changes will not increase the probability or consequences
of an accident previously evaluated.
The proposed changes to delete SRs 4.4.6.2.1.a and 4.4.6.2.1.b
and revise SR [Surveillance Requirement] 4.4.6.2.1.d will not cause
an accident to occur and will not result in any change in the
operation of associated accident mitigation equipment. The ability
to measure RCS operational leakage will not be affected. The
proposed changes will not affect the ability to mitigate any design
basis event. The design basis accidents will remain the same
postulated events described in the Millstone Unit No. 3 FSAR, and
the consequences of the design basis accidents will remain the same.
Therefore, the proposed changes will not increase the probability or
consequences of an accident previously evaluated.
The proposed change to remove SR 4.4.6.2.2.c to perform post
maintenance testing of the RCS PIVs will not cause an accident to
occur and will not result in any change in the operation of the
associated accident mitigation equipment. The proposed change will
not revise the operability requirements (e.g., valve leakage limits)
for the RCS PIVs. Proper operation of the RCS PIVs will still be
verified, as appropriate, following maintenance activities. As a
result, the design basis accidents will remain the same postulated
events described in the Millstone Unit No. 3 FSAR, and the
consequences of the design basis accidents will remain the same.
Therefore, the proposed 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.
The proposed changes do not alter the plant configuration (no
new or different type of equipment will be installed) or require any
new or unusual operator actions. The proposed changes do not alter
the way any structure, system, or component functions and do not
alter the manner in which the plant is operated. The proposed
changes do not introduce any new failure modes. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes will not reduce the margin of safety since
they have no impact on any accident analysis assumption. The
proposed changes do not decrease the scope of equipment currently
required to be operable or subject to surveillance testing, nor do
the proposed changes affect any instrument setpoints or equipment
safety functions. The effectiveness of Technical Specifications will
be maintained since the changes will not alter the operation of any
component or system, nor will the proposed
[[Page 57121]]
changes affect any safety limits or safety system settings.
Therefore, there is no 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.
Entergy Nuclear Generation Company, Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: August 22, 2001.
Description of amendment request: The proposed amendment would
change the Technical Specification (TS) Surveillance Requirement 3/
4.7.B.1.a.2 for the Standby Gas Treatment (SBGT) System by increasing
the SBGT inlet heaters minimum output testing requirement from 14 kW to
20 kW. The associated TS Bases 3/4.7.B.1 would also be revised as a
result of the proposed TS change. The proposed change is based upon the
licensee's revised design-basis calculations for the SBGT inlet heaters
and by a modification that replaces the existing SBGT system inlet
heaters with heaters of higher output capability.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. 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. The proposed TS change does not involve a significant increase
in the probability or consequences of an accident previously evaluated.
The proposed change affects only the surveillance requirement for
the SBGT inlet heaters output capability. The SBGT heaters are not the
initiators of any accidents described in the safety analysis report
(SAR). The proposed higher inlet heater output capability test is
needed to ensure that the SBGT will continue to function as currently
designed to decrease the relative humidity (RH) of the inlet air stream
to 70% RH. The higher inlet heater output capability test does not
change the consequences of an accident previously analyzed in the SAR.
Therefore, this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed TS change does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
The proposed change to the SBGT inlet heaters capacity surveillance
testing requirement is needed to continue to ensure that the SBGT will
function to decrease the RH of the inlet air stream to 70% RH, as
assumed in the current analysis. The SBGT heaters are not the
initiators of any accidents described in the SAR. The proposed change
in the surveillance testing requirement does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed TS change does not involve a significant reduction
in the margin of safety.
The proposed higher testing acceptance criteria for the inlet
heater ensures that the SBGT will continue to function as currently
designed to decrease the RH of the inlet air steam to 70% RH. The
margin of safety is unaffected by this change. Therefore, the proposed
change does not involve a significant reduction in a margin of safety.
Based on this review, it appears that the three standards of
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Attorney for licensee: 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.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of amendment request: October 17, 2001.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) actions regarding inoperable
redundant components when an Emergency Diesel Generator (EDG) becomes
inoperable. TS 3.8.1.1 would be revised to require actions based on the
TS for the inoperable redundant component(s). The proposed revision is
consistent with NUREG-1432, Rev.2, ``Standard Technical Specifications,
Combustion Engineering Plant.''
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 operation of the facility in accordance with the
proposed amendments involve a significant increase in the
probability or consequences of an accident previously evaluated?
Neither the steam driven auxiliary feedwater pump nor the EDGs
are accident initiators, but are accident mitigators. The proposed
changes to the EDG TS do not affect the operation nor availability
of the EDGs, the motor or steam driven auxiliary feedwater pumps,
nor TS required redundant features. For those conditions that would
require a unit shutdown, once the four hour completion time had
expired, the shutdown would be performed in the manner and timeframe
supported by the existing redundant feature TS. Therefore, the
probability or consequences of any accident previously evaluated
have not been significantly increased.
2. Would operation of the facility in accordance with the
proposed amendments create the possibility of a new or different
kind of accident from any accident previously evaluated?
No new failure modes are introduced by the proposed TS changes
and single failure considerations are adequately addressed by
following the established conventions of NUREG-1432. The proposed
four hour completion time from the discovery of inoperable redundant
features and an EDG takes into account the operability of the
redundant counterpart to the inoperable required feature, the
capacity and capability of the remaining AC sources, a reasonable
time for repairs, and the low probability of a DBA [design-basis
accident] occurring during this period. The TS change required
reformatting and moving the steam driven auxiliary feedwater pump
operability requirements to the redundant feature(s) actions to be
comparable with and meet the intent of the BASES requirements
contained in NUREG-1432. Without creation of a new interaction of
materials, operating configuration, or operating interfaces, there
is no possibility that the proposed changes can introduce a new or
different kind of accident.
3. Would operation of the facility in accordance with the
proposed amendments involve a significant reduction in a margin of
safety?
The margin of safety as defined in the basis for any Technical
Specification or in any licensing document has not been reduced. The
proposed changes remove the unconditional unit shutdown requirement
should an EDG be inoperable while required features on the opposite
train are inoperable. Instead, any TS required actions are
appropriately based on the inoperability of the required feature.
The proposed four hour completion time from the discovery of
inoperable redundant features and an EDG takes into account the
operability of the redundant counterpart to the inoperable required
feature, the capacity and capability of the remaining AC sources, a
reasonable time for repairs, and the low probability of
[[Page 57122]]
a DBA occurring during this period. For those conditions that would
require a unit shutdown, once the four hour completion time had
expired, the shutdown would be performed in the manner and timeframe
supported by the existing redundant feature TS. Additionally, the TS
requirements to assure that steam driven auxiliary feedwater pump
operability is considered as part of the redundant features
requirements remains and is comparable to the intent of the BASES of
STS 3.8.1. Based on the preceding discussion, FPL concludes that the
margin of safety will not be significantly reduced by operation of
the facility in accordance with the proposed amendments.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Section Chief: Richard P. Correia.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: October 17, 2001.
Description of amendment request: The proposed amendment would
revise the Technical Specification (TS) multiplier values for single-
loop operation (SLO) average planar linear heat generation rate
(APLHGR).
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 proposed APLHGR multipliers, and their use to determine the
Cycle 21 thermal limits, have been derived using NRC approved methods
and uncertainties. These methods do not change operation of the plant,
and have no effect on the probability of an accident initiating event
or transient. The purpose of the APLHGR limit is to assure that the
fuel will not exceed a peak cladding temperature (PCT) of 2200 deg.F
during a Loss of Coolant Accident [LOCA], as required by 10 CFR 50.46.
Specifying appropriate APLHGR multipliers ensures that a LOCA in SLO
will not produce a PCT any greater than the PCT produced by a LOCA in
dual loop operation. These changes ensure that the appropriate SLO
APLHGR multiplier, required for GE14 fuel, is incorporated into the
Monticello TS. These changes do not alter the method of operating the
plant.
Therefore, the proposed TS changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
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 result only from different inputs, including
use of GE14 fuel, for the Cycle 21 core reload. These methods and
uncertainties have been reviewed and approved by the NRC, and do not
involve any new or unapproved methods for operating the facility. No
new initiating events or transients result from these changes.
The single-loop operation APLHGR multiplier values are designed to
ensure that the PCT resulting from a LOCA while operating in SLO are
bounded by the PCT from a LOCA while operating in dual loop operation.
This multiplier update results from application of GE Nuclear Energy's
(GE's) current standard methodology for this analysis.
Therefore, the proposed TS changes do not create the possibility of
a new or different kind of accident, from any accident previously
evaluated.
3. The proposed amendment will not involve a significant reduction
in the margin of safety.
The APLHGR limits are set appropriately below the value where
significant fuel damage could occur in a Loss of Coolant Accident
(LOCA). Application of new SLO APLHGR multiplier values ensure that SLO
LOCA results are bounded by those for dual loop operation and thus
maintain or improve the margin of safety for LOCA analyses.
Therefore, the proposed TS changes do not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay E. Silberg, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: William D. Reckley.
PPL Susquehanna, LLC, Docket No. 50-387, Susquehanna Steam Electric
Station, Unit 1, Luzerne County, Pennsylvania
Date of amendment request: September 19, 2001.
Description of amendment request: The proposed amendment would
revise the Unit 1 reactor pressure vessel (RPV) material surveillance
program to defer the withdrawal of the second surveillance capsule for
one operating cycle. Deferral is requested to support PPL Susquehanna,
LLC's, participation in the Boiling Water Reactor Vessel and Internals
Project Integrated Surveillance 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:
A. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Pressure-temperature (P/T) limits are imposed on the reactor
coolant system to ensure that adequate safety margins against non-
ductile or rapidly propagating failure exist during normal
operation, anticipated operational occurrences, and system
hydrostatic tests. The P/T limits are related to the nil-ductility
reference temperature, RTndt. Changes in the fracture toughness
properties of the Reactor Pressure Vessel (RPV) beltline materials,
resulting from neutron irradiation and the thermal environment, are
monitored by a surveillance program in compliance with the
requirements of 10 CFR 50, Appendix H. The effect of neutron fluence
on the shift in the nil-ductility reference temperature of pressure
vessel steel is predicted by methods given in Regulatory Guide (RG)
1.99, Revision 2 and Regulatory Guide 1.190, Revision 0. The
Susquehanna SES [Steam Electric Station] Unit 1 current P/T limits
were established based on adjusted reference temperatures developed
in accordance with the procedures prescribed in RG 1.99, Revision 2.
Calculation of adjusted reference temperature by these procedures
includes a margin term to ensure upperbound values are used for the
calculation of the P/T limits. Revision of the second capsule
withdrawal schedule will not affect the P/T limits, because they
will continue to be established in accordance with NRC approved
methodology in accordance with RG 1.190 Revision 0 commitments. The
existing P/T limits are based on 32 EFPY rather than for the planned
withdrawal at 15 EFPY. This change is not related to any accidents
previously evaluated. The proposed change will not affect reactor
pressure vessel performance because no physical changes are involved
and the RPV vessel P/T limits will remain in accordance with RG
1.99, Revision 2 commitments. The proposed change will not cause the
reactor pressure vessel or
[[Page 57123]]
interfacing safety systems to be operated outside of their design or
testing limits. Also, the proposed change will not alter any
assumptions previously made in evaluating the radiological
consequences of accidents.
Therefore, this proposed amendment does not involve a
significant increase in the probability of occurrence or
consequences of an accident previously evaluated.
B. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously analyzed?
The proposed change defers the second RPV material surveillance
capsule withdrawal for one fuel cycle. This proposed change does not
involve a modification of the design of plant structures, systems,
or components. The proposed change will not impact the manner in
which the plant is operated as plant operating and testing
procedures will not be affected by the change. The proposed change
will not degrade the reliability of structures, systems, or
components important-to-safety because equipment protection features
will not be deleted or modified, equipment redundancy or
independence will not be reduced, supporting system performance will
not be downgraded, the frequency of operation of equipment
important-to-safety will not be increased, and more severe testing
of equipment important-to-safety will not be imposed. No new
accident types or failure modes will be introduced as a result of
the proposed change.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from previously
analyzed.
C. Does the proposed change involve a significant reduction in a
margin of safety?
Appendix G to 10 CFR 50 describes the conditions that require P/
T limits and provide the general bases for these limits. Until the
results from the reactor vessel surveillance program become
available, RG 1.99, Revision 2 is used to predict the amount of
neutron irradiation damage. The use of operating limits based on
these criteria, as defined by applicable regulations, codes, and
standards, provide reasonable assurance that nonductile or rapidly
propagating failure will not occur. The P/T limits are not derived
from Design Basis Accident (DBA) analyses. They are prescribed
during normal operation to avoid encountering pressure, temperature,
and temperature rate of change conditions that might cause
undetected flaws to propagate and cause nonductile failure of the
reactor coolant pressure boundary (RCPB). Since the P/T limits are
not derived from any DBA, there are no acceptance limits related to
the P/T limits. Rather, the P/T limits are acceptance limits
themselves since they preclude operation in an unanalyzed condition.
The proposed change will not affect any safety limits, limiting
safety system settings, or limiting conditions of operation. The
proposed change does not represent a change in initial conditions,
or in a system response time, or in any other parameter affecting
the course of an accident analysis supporting the Bases of any
Technical Specification. The proposed change does not involve
revision of the P/T limits, but rather a revision of the withdrawal
time for the second surveillance capsule. The current P/T limits
were established based on adjusted reference temperatures for vessel
beltline materials calculated in accordance with RG 1.99, Revision
2. P/T limits will continue to be revised, as necessary, for changes
in adjusted reference temperature due to changes in fluence when two
or three credible surveillance data sets become available. When two
or more credible surveillance data sets become available, P/T limits
will be revised as prescribed in RG 1.190, Revision 0.
Therefore, the proposed changes do not involve a significant
reduction in any margins 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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: L. Raghavan, Acting.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: July 30, 2001, as supplemented August 7,
and October 16, 2001.
Description of amendment request: The proposed amendment would
revise Technical Specification 5.5.12, ``Primary Containment Leakage
Rate Testing Program,'' to allow a one-time deferral of the Type A
containment integrated leakage rate test (ILRT) at the Susquehanna
Steam Electric Station (SSES), Units 1 and 2. The Unit 1 test would be
deferred to no later than May 3, 2007, and the Unit 2 test would be
deferred to no later than October 30, 2007, resulting in an extended
interval of 15 years for performance of the next ILRT at each unit.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
The frequency of Type A testing does not change the probability
of an event that results in core damage or vessel failure. Primary
containment is the engineered feature that contains the energy and
fission products from evaluated events. The SSES IPE [Individual
Plant Examination] documents events that lead to containment
failure. The frequency of events that lead to containment failure
does not change because it is not a function of the Type A test
interval. Containment failure is a function of loss of safety
systems that shutdown the reactor, provide adequate core cooling,
provide decay heat removal, and drywell sprays.
The consequences of the evaluated accidents are the amount of
radioactivity that is released to secondary containment and
subsequently to the public. Normally, extending a test interval
increases the probability that a Structure System or Component will
be failed. However, NUREG-1493, Performance-Based Containment Leak-
Test Program, states that calculated risks in BWR's is very
insensitive to the assumed leakage rates. The remaining testing and
inspection programs provide the same coverage as the Type A test.
These other programs will maintain containment leakage low. Any
leakage path problems will be identified and repairs will be made.
Additionally the containment is continuously monitored during power
operation. Anomalies are investigated and resolved. Thus there is a
high confidence that containment integrity will be maintained
independent of the Type A test frequency.
Therefore, this proposed amendment does not involve a
significant increase in the probability of occurrence 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 analyzed?
Primary containment is designed to contain energy and fission
products during and after an event. The SSES IPE identifies events
that lead to containment failure. Revision to the Type A test
interval does not change this list of events. There are no physical
changes being made to the plant and there are no changes to the
operation of the plant that could introduce a new failure mode
creating an accident or affecting mitigation of an accident.
Therefore, this proposed amendment does not involve a
possibility of a new or different kind of accident from any
previously analyzed.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed revision to Technical Specifications adds a one
time extension to the current interval for Type A testing. The
current level of 10 years, based on past performance, would be
extended on a one time basis to 15 years from the last Type A test.
The NUREG-1493 generic study of the effects of extending containment
leakage testing found that a 20-year interval in Type A leakage
testing resulted in an imperceptible increase in risk to the public.
NUREG-1493 found that, generically, the design containment leakage
rate contributes about 0.1% to the individual risk and that
increasing the Type A test interval would have minimal affect on
this risk since 95% of the potential leakage paths are detected by
Type B and Type C testing. Technical Specifications require that
maximum allowable primary containment leakage rate is less than 1%
primary containment air
[[Page 57124]]
weight per day. During unit startup following Type B and Type C
testing, leakage rate acceptance criteria must be less than 0.6%
primary containment air weight per day. (TS 5.5.12) Therefore, Type
B and Type C testing combined with visual inspection programs will
maintain containment leakage low.
Therefore, these changes do 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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: L. Raghavan, Acting.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: April 16, 2001, as supplemented on July
5, 2001.
Description of amendment request: The proposed Technical
Specifications (TSs) change would modify required actions and
surveillance requirements (SR) associated with the 28 Volt Direct
Current (VDC) Battery System. The proposed changes are consistent with
TS and SR requirements for the 125 VDC Battery System, and NUREG-1431,
``Standard Technical Specifications--Westinghouse Plants.''
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. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to TS limiting conditions for operation (LCOs)
and surveillance requirements (SRs) will not alter the plant's physical
configuration or the operation of the 28 VDC Battery System. As a
result, the parameters assumed in the Salem Updated Final Safety
Analysis Report (UFSAR) Design Basis Accident or Transient Analyses
remain unchanged. Therefore, the probability or consequences of an
accident previously evaluated are not increased by the proposed change.
2. Does not create the possibility of a new or different kind of
accident from any accident previously analyzed.
The proposed changes to the 28 VDC Battery System TS LCOs and SRs
do not modify the facility's design or physical configuration or change
the method by which any safety-related system performs its function.
Therefore, the proposed changes will not increase the possibility of a
new or different kind of accident from any accident previously
identified.
3. The proposed changes do not involve a significant reduction in
the margin of safety.
The proposed changes do not alter the manner in which safety limits
or limiting safety system setpoints are determined. As a result,
margins of safety are not changed. Therefore, the proposed changes do
not involve a significant reduction in any 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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: September 24, 2001.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.9, ``Refueling Operations,'' by
relocating requirements for boron concentration to the Core Operating
Limits Report (COLR). The proposed amendment will revise Limiting
Condition for Operation (LCO) 3.9.1 by stating that, while the plant is
in Mode 6, boron concentration of the Reactor Coolant System (RCS),
refueling canal, and the refueling cavity shall be maintained within
the limits specified in the COLR. LCO 3.9.1 required actions will also
be revised to reference the COLR, and associated surveillance
requirements will be changed to state that boron concentration shall be
verified to be within the limits provided in the COLR every 72 hours.
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. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Relocating the minimum required boron concentration values from the
TSs to the COLR does not change boron concentration requirements.
Specifying the required minimum boron concentration in the COLR will
continue to ensure that the proper boron concentration will be
maintained in accordance with all the assumptions of appropriate
accident analyses. Therefore, the proposed change will not involve a
significant increase in the probability or consequences of an accident
previously evaluated.
2. Does not create the possibility of a new or different kind of
accident from any accident previously analyzed.
The proposed change relocates the minimum required boron
concentration values from the TSs to the COLR. Moreover, the proposed
change does not physically change the facility, plant operations, or
the manner and frequency at which associated boron concentration
testing is conducted. Therefore, the proposed change to relocate the
required boron concentration to the COLR does not create the
possibility of a new or different kind of accident from any previously
analyzed.
3. The proposed changes do not involve a significant reduction in
the margin of safety.
Minimum boron concentration limits are established to ensure that
sufficient margins exist to prevent criticality in the RCS, refueling
canal, and the refueling cavity during refueling operations. Since the
COLR is prepared as part of each core reload safety evaluation to
ensure that current safety analysis limits are met, relocating the
minimum boron concentration from the TSs to the COLR will not reduce
safety margins. Therefore, the new proposed change to relocate the
required boron concentration to the COLR does not involve a significant
reduction in the 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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
[[Page 57125]]
South Carolina Electric & Gas Company (SCE&G), South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station,
Unit No. 1, Fairfield County, South Carolina
Date of amendment request: June 19, 2001.
Description of amendment request: South Carolina Electric & Gas
Company (SCE&G) proposes a change to the Virgil C. Summer Nuclear
Station (VCSNS) Technical Specifications (TS) Surveillance Requirements
to revise Table 3.7-1. This change will identify maximum allowable
power range neutron flux high setpoints based on the plant safety
analysis or conservatively derived values calculated in accordance with
NRC Information Notice 94-60 and Westinghouse Nuclear Safety Advisory
Letter NSAL-94-001.
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 changes to Technical Specification 3.7.1 and its
associated bases do not contribute to the initiation of any accident
previously evaluated. Supporting factors are as follows:
All NSSS components are compatible with the revised core power
limits and resulting operating conditions. Their structural
integrity is maintained during all proposed plant conditions through
compliance with the ASME code.
Other systems important to safety are not adversely impacted and
will continue to perform their design functions.
The revised core power limits and resulting operating conditions
remain within the design envelope of the plant.
Therefore, since the reactor coolant pressure boundary integrity
and system functions are not adversely impacted, the probability of
occurrence of an accident previously evaluated will be no greater
than the existing design basis of the plant. The revised method to
derive allowable power levels with inoperable main steam safety
valves results in lower High Flux Trip Setpoints. When implemented,
the revised trip setpoints ensure that secondary system pressure
will be limited to within 110% (1305 psig) of its design pressure of
1185 psig during the most severe anticipated system operational
transients. Since the ASME and regulatory limits on secondary side
overpressurization will be met, the proposed changes will not create
the potential for an increase in offsite releases or doses for any
accident. Therefore, there is no 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 changes to Technical Specification 3.7.1 and its
associated bases do not introduce any new accident initiator
mechanisms. Structural integrity of the RCS and the secondary side
is maintained during the allowed operating conditions, and ASME code
limits continue to be met during all anticipated operating
conditions. In addition, no new failure modes or limiting single
failure or new design requirements for auxiliary systems are being
introduced. Since the safety and design requirements continue to be
met and the integrity of the primary and secondary pressure boundary
is maintained, no new accident scenarios have been created.
Therefore, the types of accidents previously defined continue to
represent the credible spectrum of events to be analyzed. A new or
different kind of accident is thus not created.
3. Does this change involve a significant reduction in margin of
safety?
The proposed changes to Technical Specification 3.7.1 and its
associated bases preserve the results and conclusions of plants
safety analyses presented in the FSAR. The proposed changes address
an identified deficiency with the current Technical Specification
and, when implemented, restores the margin of safety intended.
Specifically, the proposed changes ensures overpressure ensure that
the secondary system pressure will be limited to within 110% (1305
psig) of its design pressure of 1185 psig during the most severe
anticipated system operational transient. Therefore, there is no
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Thomas G. Eppink, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Section Chief: Richard L. Emch, Jr.
Southern Nuclear Operating Company, Inc., Georgia Power Company,
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch
Nuclear Plant, Units 1 and 2, Appling County, Georgia
Date of amendment request: September 19, 2001.
Description of amendment request: The proposed amendments would
revise surveillance requirement 3.6.1.3.8 which currently requires
verification of the actuation capability of each reactor
instrumentation excess flow check valve (EFCV) every 18 months. The
proposed amendments would state that a representative sample of the
EFCVs will be tested every 18 months such that each EFCV will be tested
at least once every 10 years. The proposed amendments are consistent
with Technical Specification Task Force-334.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of a previously evaluated event?
The Excess Flow Check Valves are designed to limit the flow from
an instrument line break downstream of the check valve itself. Thus
the previously analyzed event is the instrument line break,
documented in the Unit 2 FSAR, section 15.4.13, for both units. This
proposed revision does not alter the operation or maintenance of any
instrument lines; the revision is made to reduce the surveillance
requirements for the EFCVs. This revision does nothing which
jeopardizes the integrity of the instrument lines and thus increase
the probability of a line break.
The line break analysis does not take credit for operation of
the excess flow check valves, therefore, the radiological
consequences of this event are not affected by this proposed TS
revision.
This amendment request does not affect any other previously
evaluated line or pipe break analsis.
For the above reasons, the probability of occurrence, or the
consequences of a previously evaluated event are not increased by
this proposed change.
2. Do the proposed changes create the possibility of a new type
event different from any previously evaluated?
No changes are being made to the way in which the EFCVs are
operated, or maintained; they will continue to be operated within
the conditions for which they were designed. Since no new
operational modes are proposed, no new failure modes are introduced.
Furthermore, no changes to any systems designed for the
prevention of transients or accidents are being made as a result of
this proposed Technical Specification change.
For the above reasons, this proposed change does not introduce
the possibility of a different type event from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
The reactor coolant pressure boundary line break analysis
documented in Unit 2 FSAR section 15.4.13 does not assume credit for
the EFCVs. Additionally, the failure rate of the Unit 1 and 2 EFCVs
has been small, as verified by the failure rate analysis done for
this proposed revision. Accordingly, reducing the frequency of the
surveillance is justified and will not significantly reduce the
margin of safety with respect to EFCV failure.
Additionally, General Electric has performed a generic
radiological evaluation of an instrument line break, with EFCV
failure, which concluded that the dose
[[Page 57126]]
consequences would not exceed 10 CFR 100 guidelines. This analysis
is documented in NEDO-32977-A, ``Excess Flow Check Valve
relaxation'', a report commissioned by the Boiling Water Reactors
Owners' Group (BWROG). Because the Hatch EFCV design is similar to
the EFCV designs assumed in the NEDO, it is reasonable to conclude
that the results of this generic analysis are bounding for Plant
Hatch.
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: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC
20037.
NRC Section Chief: Richard J. Laufer, Acting.
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 amendment request: August 10, 2001.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3.9.1, ``Refueling Equipment
Interlocks,'' to provide alternative actions when the refueling
equipment interlocks are inoperable.
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 does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The operation of refueling interlocks is explicitly assumed in
the analyses of the control rod removal error and fuel loading error
during refueling. Inadvertent criticality is prevented during the
loading of fuel provided all control rods are fully inserted. The
refueling interlocks accomplish this by preventing the loading of
fuel into the core with any control rod withdrawn, or by preventing
withdrawal of a rod from the core during fuel loading. Under
existing TS when the refueling interlocks are inoperable, the
current method of preventing fuel loading with control rods
withdrawn is to prevent fuel movement. An alternate method to ensure
that fuel is not loaded into a cell with a control rod withdrawn is
to prevent control rods from being withdrawn and to verify that all
control rods are fully inserted. The proposed TS Required Actions
will require that a control rod block be placed in effect, thereby
ensuring that control rods are not subsequently inappropriately
withdrawn, and that all required control rods be verified to be
fully inserted. This verification is in addition to the requirements
to periodically verify control rod position by other TS
requirements.
The proposed actions will ensure that control rods are not
withdrawn and cannot be inappropriately withdrawn, because a control
rod withdrawal block is in place. Like the current TS requirements,
the proposed actions will ensure that unacceptable operations are
blocked. Hence, the proposed additional Required Actions provide an
equivalent level of assurance that fuel will not be loaded into a
core cell with a control rod withdrawn as does the current TS
Required Action. Therefore, the proposed change does not
significantly increase the probability or consequences of an
accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The change in the TS requirements does not involve a change in
plant design or to the analyzed condition of the reactor core during
refueling. The proposed new Required Actions will ensure that
control rods are not withdrawn and cannot be inappropriately
withdrawn, because a block to control rod withdrawal is in place.
Therefore, no new failure modes are introduced, and the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
As discussed in the Bases for the affected TS requirements,
inadvertent criticality is prevented during the loading of fuel
provided all control rods are fully inserted during the fuel
insertion. The refueling interlocks function to support the
refueling procedures by preventing control rod withdrawal during
fuel movement and the inadvertent loading of fuel when a control rod
is withdrawn. The proposed change will allow the refueling
interlocks to be inoperable and fuel movement to continue only if a
control rod withdrawal block is in effect and all control rods are
verified to be fully inserted. These proposed Required Actions
provide an equivalent level of protection as the refueling
interlocks by preventing a configuration that could lead to an
inadvertent criticality event. The refueling procedures will
continue to be supported by the proposed Required Actions because
control rods cannot be withdrawn and as a result, fuel cannot be
inadvertently loaded when a control rod is withdrawn. Therefore, the
proposed changes do 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry
Nuclear Plant, Units 2 and 3, Limestone County, Alabama
Date of amendment request: August 17, 2001.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Table 3.3.1.1-1, ``Reactor
Protection System [RPS] Instrumentation,'' to remove one RPS function
and modify another.
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:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Modifications to the Scram Discharge Instrument Volume (SDIV)
System are being implemented to ensure that the SDIV high water
level instrumentation will respond adequately to provide redundant,
diverse trip functions for a Scram Discharge Volume (SDV) inleakage
event. Since the scram function will be successfully performed, the
removal of the low scram pilot air header pressure trip function
does not involve a significant increase in the probability or
consequences of any accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The design criteria for the Scram Discharge System is contained
in the Safety Evaluation Report on the BWR Scram Discharge System,
which was transmitted by NRC letter dated December 9, 1980, to all
BWR licensees. Modifications to the SDV System have been evaluated
to demonstrate that the high water level instrumentation in the SDIV
will respond adequately to provide the required trip function. No
new system failure modes are created as a result of removing the low
scram pilot air header trip, since the redundant and diverse SDIV
high water level instruments will initiate a successful reactor
scram. Therefore, the removal of the low scram pilot air header trip
function does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The water level in the SDIV is monitored by both resistance-
temperature type detectors and float switches. Redundancy and
diversity in the instrumentation that initiates the
[[Page 57127]]
scram signal is maintained even with the removal of the low scram
pilot air header pressure trip function. Modifications to the SDIV
System have been evaluated to demonstrate that the high water level
instrumentation will respond adequately to provide the required trip
function for an inleakage event. Therefore, the proposed amendment
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant,
Unit 2, Hamilton County, Tennessee
Date of application for amendments: October 9, 2001 (TS 01-10).
Brief description of amendments: The proposed amendment would
change the Sequoyah (SQN) Unit 2 Operating License Technical
Specifications (TSs), specifically TS 6.8.4.h, ``Containment Leakage
Rate Testing Program,'' to allow a one-time 5-year extension to the
current 10-year test interval for the containment performance-based
leakage rate test program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Tennessee Valley
Authority (TVA), the licensee, has provided its analysis of the issue
of no significant hazards consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed extension to Type A testing does not increase the
probability of an accident previously evaluated since the change is
not a modification to plant systems, nor a change to plant operation
that could initiate an accident.
TVA performed an evaluation of the risk significance for the
proposed increase to the Sequoyah Unit 2 Type A test frequency. The
results of the TVA evaluation indicate that the increase in Large
Early Release Frequency (LERF) remains below the level of risk
significance defined in NRC Regulatory Guide (RG) 1.174, ``An
Approach for Using Probabilistic Risk Assessment In Risk-Informed
Decisions On Plant-Specific Changes to the Licensing Basis.'' TVA's
evaluation indicates that the increase in frequency for all releases
(small, large, early and late) and the increase in radiation dose to
the population is non-risk significant (3.5E-7/reactor year and 7.72
person-rem, respectively).
The proposed test interval extension does not involve a
significant increase in the consequences of an accident because
research documented in NUREG-1493 determined that generically, very
few potential containment leakage paths fail to be identified by
Type A tests. An analysis of 144 Type A test results, including 23
failures, found that no failures were due to containment liner
breach. The NUREG concluded that reducing the Type A test frequency
to once per 20 years would lead to an imperceptible increase in
risk. Furthermore, the NUREG concluded that Type B and C testing
provides assurance that containment leakage from penetration leak
paths (i.e., valves, flanges, containment air-locks) identify any
leakage that would otherwise be detected by the Type A tests.
In addition to the NUREG conclusions, TVA's American Society of
Mechanical Engineers (ASME) IWE program performs containment
inspections periodically in order to detect evidence of degradation
that may affect either the containment structural integrity or leak
tightness. Accordingly, TVA's proposed extension of the Type A test
interval does not [significantly] increase the probability or
consequences of an accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change to extend the Type A test interval does not
create the possibility of a new or different type of accident since
there are no physical changes made to the plant. There are no
changes to the operation of the plant that would introduce a new
failure mode creating the possibility of a new or different kind of
accident.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change to extend the Type A test interval will not
significantly reduce the margin of safety. A generic study
documented in NUREG-1493 indicates that extending the Type A leak
test interval to 20 years would result in an imperceptible increase
in risk to the public. The NUREG also found that, generically, the
containment leakage rate contributes a very small amount to the
individual risk and that the decrease in the Type A test frequency
would have a minimal affect on risk because most potential leakage
paths are detected by Type C testing.
Previous Type A leakage tests conducted on Sequoyah Unit 2
indicate that leakage from Unit 2 containment has been less than the
10 CFR 50 Appendix J leakage limit of 1.0 La. A review of
previous Unit 2 Type A test results indicate at least a 10 percent
margin exists below the 1.0 La leakage limit. These test
results provide assurance that the proposed extension to the Type A
test interval would not significantly reduce the margin of safety.
The NRC 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 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Richard P. Correia.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, 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/NRC/ADAMS/index.html. If you do not have access to ADAMS or if there
are problems in accessing the
[[Page 57128]]
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].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1, Dauphin County, Pennsylvania
Date of application for amendment: May 31, 2001, as supplemented
September 14, 2001.
Brief description of amendment: The amendment revised the TMI-1
Technical Specifications (TSs) to incorporate Cycle 14 specific limits
for the variable low reactor coolant system pressure-temperature core
protection safety limits. These changes are reflected in revisions to
Figures 2.1-1 and 2.1-3 of the TSs and the related Bases.
Date of issuance: October 23, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 238.
Facility Operating License No. DPR-50.: Amendment revised the TSs.
Date of initial notice in Federal Register: July 11, 2001 (66 FR
36337).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 23, 2001.
No significant hazards consideration comments received: No.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of application for amendment: October 30, 2000, as
supplemented by letter dated September 13, 2001.
Brief description of amendment: The amendment modified the Final
Safety Analysis Report (FSAR) to reflect analysis of a HI-STORM 100
spent fuel cask system, spent fuel pool description and crane
operations.
Date of issuance: October 26, 2001.
Effective date: October 26, 2001, and shall be implemented in the
next periodic update to the FSAR in accordance with 10 CFR 50.71(e).
Amendment No.: 174.
Facility Operating License No. NPF-21: The amendment revised the
Final Safety Analysis Report.
Date of initial notice in Federal Register: March 21, 2001 (66 FR
15918). The September 13, 2001, supplemental letter provided additional
clarifying information, did not expand the scope of the original
Federal Register notice, and did not change the staff's original
proposed no significant hazards consideration determination. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated October 26, 2001.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: August 23, 2001, as supplemented
by letter dated September 25, 2001.
Brief description of amendment: The amendment revised the Technical
Specifications to eliminate the requirement to move control element
assembly #43 for the remainder of Cycle 15.
Date of issuance: October 22, 2001.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment No.: 235.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 5, 2001 (66
FR 46478). The September 25, 2001, supplemental letter provided
clarifying information that was within the scope of the original
Federal Register notice and did not change the staff's initial no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 22, 2001.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 18, 2001.
Brief description of amendment: The amendment changes Technical
Specifications (TS) Definitions 1.12 and 1.25, the effect of which will
be to allow either an allocated or a measured response time to be
utilized for the sensors in the Reactor Protective System and
Engineered Safety Features Actuation System instrument loops.
Date of issuance: October 29, 2001.
Effective date: As of the date of issuance and shall be implemented
60 days from the date of issuance.
Amendment No.: 175.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 5, 2001 (66
FR 46479).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 29, 2001.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: October 24, 2000, as
supplemented by letters dated June 18 and August 21, 2001.
Brief description of amendment: The amendment revises Technical
Specification 3.8.3 regarding the lube oil inventories for the Grand
Gulf Nuclear Station, Unit 1, Divisions I, II, and III emergency diesel
generators (EDGs), and will result in additional margins for lube oil
availability to provide for EDG operability for seven days following a
postulated design basis accident.
Date of issuance: October 23, 2001.
Effective date: As of the date of issuance and shall be implemented
within 90 days of issuance.
Amendment No: 149.
Facility Operating License No. NPF-29: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: January 24, 2001 (66 FR
7680).
The June 18 and August 21, 2001 supplemental letters did not change
the scope of the original Federal Register notice or the initial no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 23, 2001.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois, Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will
County, Illinois
Date of application for amendments: October 1, 2001, as
supplemented by letters dated October 9, 2001, and October 18, 2001.
The supplemental letters provided clarifying information only and did
not change the original proposed no significant hazards determination.
Brief description of amendments: The amendments revise Byron and
Braidwood technical specifications (TS) surveillance requirement (SR)
3.7.2.1 and SR 3.7.2.2 to add a note stating that these surveillances
are not required to be met until the first startup after September 27,
2001. This change is
[[Page 57129]]
applicable to Byron Station Units 1 and 2, and Braidwood Unit 2 only.
This change is not applicable to Braidwood Station, Unit 1, due to the
recent restart of the unit after the refueling outage.
Date of issuance: November 1, 2001.
Effective date: November 1, 2001.
Amendment Nos.: 124, 124, 119, and 119.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: October 23, 2001 (66 FR
53643).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 1, 2001.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: December 2, 2000, as
supplemented by letters dated September 4 and September 28, 2001.
Brief description of amendment: This amendment increases the spent
fuel pool (SFP) storage capability, as a result of the SFP re-racking
project, from the current capacity of 735 fuel assemblies to a new
capacity of 1624 fuel assemblies. The amendment also approves
additional temporary storage of up to 90 fuel assemblies in the fuel
transfer pit to support a complete re-racking of the SFP. The increase
in SFP storage capacity will provide a full core offload capability
during the plant's Cycle 13 operation and enable the Davis-Besse
facility to meet its storage needs through April 22, 2017, which is the
expiration date for the current operating license.
Date of issuance: October 19, 2001.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 247.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 6, 2001 (66
FR 46656).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 19, 2001.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346 Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: April 4, 2001.
Brief description of amendments: This license amendment request:
Deletes Technical Specification (TS) 1.7, Definitions-Reportable Event,
and TS 6.6, Reportable Event--Action; Revise TS 6.5.3, Technical Review
and Control--Activities, and TS Bases 4.0.3, Applicability.
Date of issuance: November 2, 2001.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 248.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 12, 2001 (66
FR 31708).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 2, 2001.
No significant hazards consideration comments received: No.
Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie
Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: June 22, 2001, as supplemented
August 24, 2001.
Brief description of amendment: Revised Technical Specifications to
allow the containment equipment door and airlock doors to be open
during core alterations and fuel movement under administrative
controls.
Date of Issuance: October 22, 2001.
Effective Date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment No.: 120.
Facility Operating License No. NPF-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 19, 2001 (66
FR 48287).
The August 24, 2001, supplement did not affect the original
proposed no significant hazards determination, or expand the scope of
the request as noticed in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 22, 2001.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: October 24, 2000, as
supplemented June 29, 2001.
Brief description of amendments: The amendments would approve
changes to the updated final safety analysis report to incorporate a
supplemental methodology into the analysis of steam generator overfill
following a steam generator tube rupture.
Date of issuance: October 24, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 256 and 239.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 24, 2001 (66 FR
7682). The supplemental letter contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice. The Commission's related evaluation of the amendments
is contained in a Safety Evaluation dated October 24, 2001.
No significant hazards consideration comments received: No.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of application for amendments: June 12, 2000, as supplemented
by letters dated November 7, 2000, June 19, and August 17, 2001.
Brief description of amendments: The amendments revised the
technical specifications to change the standard by which you test
charcoal used in engineered safeguard features systems to American
Society for Testing and Materials D3808-1989. These revisions are made
in accordance with Generic Letter 99-02, ``Laboratory Testing of
Nuclear-grade Activated Charcoal.''
Date of issuance: October 24, 2001.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 257 and 240.
Facility Operating License Nos. DPR-58 and DPR-74: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 23, 2000 (65 FR
51356). The supplemental letters contained clarifying information and
did not change the initial no significant hazards consideration
determination
[[Page 57130]]
and did not expand the scope of the original Federal Register notice.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated October 24, 2001.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: February 28, 2001, as supplemented by
letters dated September 14, 18, and 27, 2001. The letters dated
September 14, 18, and 27, 2001, provided clarifying information, and
did not alter the NRC staff's conclusions regarding finding of no
significant hazards consideration.
Brief description of amendment: The amendment evaluates the
licensee's revised calculation methodology for assessment of
consequences of design basis accidents, and revises Technical
Specifications.
Date of issuance: October 23, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 187.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 19, 2001 (66
FR 48288).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 23, 2001.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: April 12, 2001.
Brief description of amendment: The Amendment revises the Technical
Specifications Bases Control Program to incorporate revisions to 10 CFR
50.59.
Date of issuance: October 25, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 188.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 19, 2001 (66
FR 48289).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 25, 2001.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: April 12, 2001.
Brief description of amendment: The amendment revises the Technical
Specifications surveillance test requirement SR 3.6.1.3.8, for excess
flow check valves (EFCVs), to relax the 18-month EFCV surveillance
frequency by limiting the number of tests to a ``representative
sample'' every 18 months, such that each EFCV will be tested at least
once every 10 years.
Date of issuance: October 26, 2001.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No.: 189.
Facility Operating License No. DPR-46: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 19, 2001 (66
FR 48289).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 26, 2001.
No significant hazards consideration comments received: No.
North Atlantic Energy Service Corporation, et al., Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: June 12, 2001.
Description of amendment request: The amendment changes Technical
Specification (TS) 4.4.10 to incorporate alternative reactor coolant
pump flywheel inspections and makes administrative wording changes to
TSs 6.4.1.7.b, 6.4.2.2.d, and 6.4.2.3.
Date of issuance: October 22, 2001.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment No.: 79.
Facility Operating License No. NPF-86: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 25, 2001 (66 FR
38764).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 22, 2001.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: May 18, 2001, as supplemented
October 10, 2001.
Brief description of amendment: The amendment (1) deletes a
redundant requirement for valving out control rod drives, (2) revises
control rod accumulator operability requirements, (3) adds the option
to hydraulically isolate control rod drives, and (4) corrects an
inconsistency describing when source range monitors are required to be
operable during core monitoring.
Date of issuance: October 26, 2001.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 123.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 12, 2001 (66 FR
31711).
The supplement provided clarifying information to the application
that was within the scope of the original Federal Register notice and
did not change the staff's initial proposed no significant hazards
considerations determination. The Commission's related evaluation of
the amendment is contained in a Safety Evaluation dated October 26,
2001.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of application for amendment: August 15, 2001.
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) to (1) reflect the replacement of Monticello's
licensed operator initial and requalification training programs with an
accredited systems-approach-to-training program and (2) relocate the
existing TS requirements for procedures, records, and reviews to the
Operational Quality Assurance Plan.
Date of issuance: October 30, 2001.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 124.
Facility Operating License No. DPR-22: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 19, 2001 (66
FR 48290).
[[Page 57131]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 30, 2001.
No significant hazards consideration comments received: No.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of application for amendments: August 31, 2001.
Brief description of amendments: The amendments extend the
implementation date for Amendment No. 184 for Unit 1 and Amendment No.
158 for Unit 2 from November 1, 2001, to November 1, 2003. Amendment
Nos. 184 and 158 approved technical specification changes to
incorporate requirements related to oscillation power range monitoring
(OPRM) instrumentation. The implementation date extension is needed to
provide additional time to address software deficiencies with the OPRM
system identified in a June 29, 2001, General Electric report filed
pursuant to part 21 of Title 10 of the Code of Federal Regulations.
Date of issuance: October 29, 2001.
Effective date: As of date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 196 and 172.
Facility Operating License Nos. NPF-14 and NPF-22: The amendments
revised the license.
Date of initial notice in Federal Register: September 19, 2001 (66
FR 48291).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated October 29, 2001.
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: October 30, 2000, as
supplemented by letters dated May 7, June 13, 2001, and by internet
memoranda dated June 28, July 3, July 23, and October 16, 2001.
Brief description of amendments: Amendment Application No. 217 is a
request to revise the San Onofre Nuclear Generating Station, Unit 1
(SONGS 1) operating license and technical specifications to remove
certain requirements that have been determined to be unnecessary and
modify requirements to provide flexibility during the decommissioning
of SONGS 1. This change removes the need to perform activities that are
not providing a benefit to safely maintain the spent fuel in the spent
fuel pool. This change also provides some flexibility in the operation
of the spent fuel pool during the decommissioning of SONGS 1.
Date of issuance: October 30, 2001.
Effective date: October 30, 2001, to be implemented within 30 days
of issuance.
Amendment No.: Unit 1-160.
Facility Operating License No. DPR-13: The amendment revised the
Operating License and the Technical Specifications.
Date of initial notice in Federal Register: December 13, 2000 (65
FR 77924).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated October 30, 2001.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: August 6, 2001 (TS 01-05).
Brief description of amendments: The amendments revised the SQN
Unit 1 and 2 Technical Specifications (TSs) by changing the
surveillance requirements for verifying that containment isolation
valves to be closed. More specifically, valves in high radiation areas
may be verified by administrative means. In addition, valves which are
locked sealed or otherwise secured do not need to be reverified closed
and are eliminated from the scope of the surveillance.
Date of issuance: October 24, 2001.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment Nos.: 271 and 260.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the TSs.
Date of initial notice in Federal Register: August 22, 2001 (66 FR
44177). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 24, 2001.
No significant hazards consideration comments received: No.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of application for amendment: April 23, 2001.
Brief description of amendment: The amendment updates the license
by deleting obsolete information, correcting errors, and making
administrative changes to enhance the context and provide consistency.
Date of Issuance: October 22, 2001.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 206.
Facility Operating License No. DPR-28: Amendment revised the
License.
Date of initial notice in Federal Register: May 30, 2001 (66 FR
29363).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated October 22, 2001.
No significant hazards consideration comments received: No.
Note: The publication date for this notice will change from
every other Wednesday to every other Tuesday, effective January 8,
2002. The notice will contain the same information and will continue
to be published biweekly.
Dated at Rockville, Maryland, this 6th day of November 2001.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 01-28399 Filed 11-13-01; 8:45 am]
BILLING CODE 7590-01-P