[Federal Register Volume 64, Number 86 (Wednesday, May 5, 1999)]
[Notices]
[Pages 24192-24211]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-11119]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from April 10, 1999, through April 23, 1999. The
last biweekly notice was published on April 21, 1999 (64 FR 19554).
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 Administration 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, the Gelman Building, 2120 L Street, NW, Washington, DC. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By June 4, 1999, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW, Washington, DC and at the local public
document room for the particular facility involved. 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
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amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner who fails
to file such a supplement which satisfies these requirements with
respect to at least one contention will not be permitted to participate
as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
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, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: April 12, 1999.
Description of amendment request: The proposed one-time technical
specification (TS) change, effective through September 30, 1999,
provides a Required Action and Completion Time for the Ultimate Heat
Sink (UHS) in the event that service water temperature exceeds the
current 95 deg.F surveillance limit. It involves an allowance to
continue operation for a period of 8 hours with the UHS at a
temperature greater than the temperature limits provided in TS Limiting
Condition of Operation 3.7.8, ``Ultimate Heat Sink (UHS)'' and provides
an upper UHS temperature limit beyond which plant shutdown is required.
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:
Carolina Power & Light (CP&L) Company has evaluated the proposed
Technical Specification change and has concluded that it does not
involve a significant hazards consideration. The conclusion is in
accordance with the criteria set forth in 10 CFR 50.92. The bases
for the conclusion that the proposed change does not involve a
significant hazards consideration are discussed below.
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change does not involve any physical alteration of
plant systems, structures or components. The proposed change allows
plant operation for a short period of time when the service water
temperature exceeds 95 deg.F, requires an hourly surveillance when
service water temperature exceeds 95 deg.F, provides an upper UHS
temperature limit beyond which a plant shutdown is required, and
specifies an expiration date beyond which the current requirements
are restored. If the service water temperature is restored within
the allowed time, a plant shutdown is not required. This minimizes
plant transients, which reduces the probability of a reactor trip
and the resulting challenges to mitigating systems. A service water
temperature of up to 99 deg.F does not increase the failure rate of
systems, structures or components because the systems, structures,
and components are designed for higher temperatures than at which
they operate.
The Service Water (SW) System temperature is not assumed to be
an initiating condition of any accident evaluated in the safety
analysis report. Therefore, the allowance of a limited time for
service water temperature to be in excess of 95 deg.F does not
involve an increase in the probability of an accident previously
evaluated in the safety analysis report (SAR). The SW System
supports operability of safety related systems used to mitigate the
consequences of an accident. The service water temperature is not
expected to increase significantly beyond 95 deg.F due to the
limited time allowed by the proposed change in conjunction with the
generally slow rate of temperature increase experienced from thermal
changes in Lake Robinson. The capability of components to perform
their safety related function is not affected up to a service water
temperature of 99 deg.F with the exception of the Containment Air
Recirculation Fan Coolers. The heat removal capacity of the
Containment Air Recirculation Fan Coolers is not expected to be
significantly reduced by a small increase in service water
temperature. If heat removal is not significantly reduced,
containment pressure and leakage will not be significantly
increased, and the doses from containment leakage will not be
significantly increased. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated in the SAR.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not involve any physical alteration of
plant systems, structures or components. A service water temperature
of up to 99 deg.F does not introduce new failure mechanisms of
systems, structures or components not already considered in the SAR
because the systems, structures, and components are designed for
higher temperatures than at which they operate. Therefore, the
possibility of a new or different kind of accident from any accident
previously evaluated is not created.
3. Does this change involve a significant reduction in a margin
of safety?
The proposed change will allow a small increase in service water
temperature above the design basis limit for the SW System and delay
by 8 hours the requirement to shutdown the plant when the service
water system design limit is exceeded. There are design margins
associated with systems, structures and components that are cooled
by the service water system that are affected. The capability of
components to perform their safety related function is not affected
up to a service water temperature [of] 99 deg.F with the exception
of the Containment Air Recirculation Fan Coolers. The Containment
Air Recirculation Fan Coolers remove heat from containment to
mitigate containment pressure and temperature following a MSLB [main
steamline break] inside containment or a Large Break LOCA [loss-of-
coolant accident] inside containment. An increase in service water
temperature in excess of the design limit due to hot weather
conditions is expected to be small due to the limited time allowed
by the proposed change in conjunction with the generally slow rate
of temperature increase experienced from thermal changes in Lake
Robinson. Therefore, the effect on the Containment Air
[[Page 24194]]
Recirculation Fan Coolers' heat removal capacity and the resulting
containment pressure and temperature is expected to be small.
Therefore, there is no significant reduction in margin of safety
associated with this change.
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.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Attorney for licensee: William D. Johnson, Vice President and
Corporate Secretary, Carolina Power & Light Company, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Sheri R. Peterson.
Commonwealth Edison Company, Docket No. 50-254, Quad Cities Nuclear
Power Station, Units 1, Rock Island County, Illinois
Date of amendment request: March 30, 1999.
Description of amendment request: The amendment would revise the
Quad Cities Nuclear Power Station, Unit 1 Technical Specifications (TS)
by changing the Surveillance Requirements (SR) 4.6.E.2 to allow a one-
time extension of the 18-month requirement to pressure set test or
replace one half of the Main Steam Safety Valves (MSSVs) to an interval
of 24 months.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes request a one-time change to the
surveillance requirement for the MSSVs and Target Rock S/RV [Safety
Relief Valve]. The surveillance interval between MSSVs and Target
Rock S/RV testing is not a precursor assumed in any previously
analyzed accident. Therefore, the probability of a previously
evaluated accident has not been increased.
The proposed extension is consistent with the ASME Code
requirement to test 20% of the sample population every 24 months
with all of the valves in the sample group being tested every 60
months. The proposed changes are also consistent with NUREG 1433,
Revision 1, and do not adversely affect existing plant safety
margins or the reliability of the equipment assumed to operate in
the safety analysis. Operating experience and excellent materiel
condition of the MSSVs and Target Rock S/RV support the expectation
that they will continue to perform their intended function.
Therefore, the consequences of a previously evaluated accident have
not been increased.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No new equipment is required, nor will the MSSVs and Target Rock
S/RV be operated in a different manner during the period of the
extended surveillance interval. The proposed changes are consistent
with NUREG 1433, Revision 1, requirements for safety valve
surveillance intervals as well as the ASME Code requirements for
testing safety valves. Operating experience and superior materiel
condition of the MSSVs and Target Rock S/RV support the expectation
that they will continue to perform their intended function.
Therefore, the possibility of a new or different accident has not
been increased.
Does the change involve a significant reduction in a margin of
safety?
The proposed amendment represents an extension to the current TS
SRs that would otherwise be provided generically by the ASME Code.
The proposed changes are also consistent with NUREG-1433, Revision
1, and do not adversely affect existing plant safety margins or the
reliability of the equipment assumed to operate in the safety
analysis. The proposed changes have been evaluated and found to be
acceptable for use at Quad Cities Nuclear Power Station based on
system safety analysis requirements and operational performance. The
MSSVs and Target Rock S/RV provisions continue to be adequately
maintained during plant operation. The proposed changes to the MSSVs
and Target Rock S/RV surveillance interval do not significantly
reduce existing plant safety margins since excellent materiel
condition and acceptable surveillance test results support the
expectation that no significant degradation will occur over the
extended interval.
The proposed changes are based on NRC accepted provisions at
other operating plants that are applicable at Quad Cities Nuclear
Power Station and maintain necessary levels of system or component
reliability.
The proposed amendment for Quad Cities Nuclear Power Station
will not reduce the availability of systems required to mitigate
accident conditions.
Therefore, these 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of amendment request: March 30, 1999.
Description of amendment request: This amendment request proposes
to change the Technical Specifications (TSs) to allow an alternate
methodology for quantifying Reactor Coolant System (RCS) leakage when
the normal RCS leakage detection system is 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:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The current Technical Specifications require a periodic
measurement of RCS leakage. The normal method for quantifying RCS
leakage is to use the DWFDS [Drywell Floor Drain Sump] and DWEDS
[Drywell Equipment Drain Sump] flow totalizers. The proposed TS
change would allow an alternate method for quantifying RCS leakage
when a flow totalizer is not available. The proposed change has no
impact on the frequency for monitoring RCS leakage and would only be
used for a maximum of 30 days while the normal leakage monitoring
system is being restored to an operable condition. The alternate
methodology for quantifying leakage has a measurement sensitivity
that is consistent with the normal method. The proposed change does
not impact any system structure or component used to mitigate the
consequences of an accident and there will be no change in the types
or significant increase in the amounts of any effluents released
offsite.
Therefore this proposed amendment does 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 change involves no physical modifications to any
system, structure or component used to mitigate the consequences of
an accident. The operation of the DWEDS and DWFDS are not being
altered in any way that could affect their ability to function
during an accident condition.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The current TS's require a periodic measurement of RCS leakage.
The normal method for quantifying RCS leakage is to use
[[Page 24195]]
the DWFDS and DWEDS flow totalizers. The proposed technical
specifications change would allow an alternate method for
quantifying RCS leakage when a flow totalizer is inoperable. The
proposed change has no impact on the frequency for monitoring RCS
leakage and would only be used for a maximum of 30-days while the
normal leakage monitoring system is being restored to an operable
condition. The proposed alternate methodology for quantifying
leakage has a measurement sensitivity that is consistent with the
normal method.
Therefore, these 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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units and, Rock Island County, Illinois
Date of amendment request: March 30, 1999.
Description of amendment request: This amendment request proposes
to revise license conditions in each of the respective Operating
Licenses to delete those license conditions that no longer apply, make
an editorial change in the Unit 1 license, and provide clarifying
information regarding the license condition concerning equalizer valve
restrictions.
Basis for proposed no significant hazards consideration
determination: As required by 10 FR 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 initial conditions and methodologies used in the accident
analyses remain unchanged. The proposed changes do not change or
alter the design assumptions for the systems or components used to
mitigate the consequences of an accident. Therefore, accident
analyses results are not impacted.
The proposed changes delete various license conditions that have
been completed, make editorial changes, and provide clarifying
information. The changes are administrative. No physical or
operational changes to the facility will result from the proposed
changes.
Therefore, this proposed amendment does 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 do not affect the design or operation of
any system, structure, or component in the plant. The safety
functions of the related structures, systems, or components are not
changed in any manner, nor is the reliability of any structures ,
systems, or component reduced. The changes do not affect the manner
by which the facility is operated and do not change any facility
design feature, structure, system, or component. No new or different
type of equipment will be installed.
The proposed changes delete various license conditions that have
been completed, make editorial changes, and provide clarifying
information. The changes are administrative. No physical or
operational changes to the facility will result from the proposed
changes.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
Does the change involve a significant reduction in the margin of
safety for the following reasons:
The proposed changes are administrative in nature and have no
impact on the margin of safety of any Technical Specification. There
is no impact on safety limits or limiting safety system settings.
The changes do not affect any plant safety parameters or setpoints.
The proposed changes delete various license conditions that have
been completed, make editorial changes, and provide clarifying
information. No physical or operational changes to the facility will
result from the proposed changes.
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
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice
President and General Counsel, Commonwealth Edison Company, P.O. Box
767, Chicago, Illinois 60690-0767.
NRC Section Chief: Anthony J. Mendiola.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: March 25, 1999.
Description of amendment request: The amendments would revise the
Technical Specifications (TS) to redefine the ``trip setpoint'' in a
number of locations as the ``nominal trip setpoint.'' The current
definition results in upper-or lower-bound numerical values not to be
exceeded for setpoints. This proposed new definition would permit the
setpoints to be set within a tolerance range around the number
specified in various tables. The TS locations affected are: Table
3.3.1-1, ``Reactor Trip System Instrumentation;'' Table 3.3.2-1,
``Engineered Safety Feature Actuation Instrumentation;'' Surveillance
Requirement 3.3.5.2; Table 3.3.6-1, ``Containment Purge and Exhaust
Isolation Instrumentation;'' and Limiting Condition of Operation (LCO)
3.4.12. Sections of the associated TS Bases document would also be
revised to reflect the TS 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, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed changes are consistent with the current
licensing basis for Catawba Nuclear Station, the setpoint
methodology used to develop the Trip Setpoints, the Catawba Safety
Analyses, and current station calibration procedures and practices.
The Reactor Trip System and Engineered Safety Features Actuation
System are not accident initiating systems; they are accident
mitigating systems. Therefore, these proposed changes will have no
impact on any accident probabilities. Accident consequences will not
be affected, as no changes are being made to the plant which will
involve a reduction in reliability of these systems. Consequently,
any previous evaluations associated with accidents will not be
affected by these changes.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. The proposed changes are consistent with the current
licensing basis for Catawba Nuclear Station, the setpoint
methodology used to develop the Trip Setpoints, the Catawba Safety
Analyses, and current station calibration procedures and practices.
No changes are being made to actual plant hardware which will result
in any new accident causal mechanisms. Also, no changes are being
made to the way in which the plant is being operated. Therefore, no
[[Page 24196]]
new accident causal mechanisms will be generated. Consequently,
plant accident analyses will not be affected by these changes.
3. Does this change involve a significant reduction in a margin
of safety?
No. The proposed changes are consistent with the current
licensing basis for Catawba Nuclear Station, the setpoint
methodology used to develop the Trip Setpoints, the Catawba Safety
Analyses, and current station calibration procedures and practices.
Margin of safety is related to the confidence in the ability of the
fission product barriers to perform their design functions during
and following accident conditions. These barriers include the fuel
cladding, the reactor coolant system, and the containment system.
The performance of these barriers will not be degraded by the
proposed changes. Consequently, plant safety analyses will not be
affected by these changes.
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.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Section Chief: Richard L. Emch, Jr.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: April 9, 1999.
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TSs) to add Limiting Condition for
Operation (LCO) 3.0.6 and its associated bases. This change would allow
equipment that has been removed from service or declared inoperable in
compliance with the TS Action statement to be returned to service under
administrative controls solely to perform testing required to
demonstrate its operability or the operability of other equipment. The
proposed change is consistent with TS 3.0.5 as discussed in NUREG-1432,
Revision 1, ``Standard Technical Specifications for Combustion
Engineering Plants.'' TS 3.0.2 would also be modified to reflect that
TS 3.0.6 is an exception to TS 3.0.2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. 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 change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed change would allow an orderly return to service of
inoperable equipment. This change does not alter the functional
characteristics of any plant component and does not allow any new modes
of operation of any component. The accident mitigation features of the
plant are not affected by the proposed amendment request. Therefore,
this proposed amendment would not result in a significant change in the
types or significant increase in the amounts of any effluents that may
be released off site. No modifications to the plant have been proposed
due to this amendment request. The proposed change would permit
equipment removed from service to comply with required actions to be
returned to service under administrative controls to verify the
operability of the equipment being returned to service or of other
related equipment. Although returning inoperable equipment to service
for testing may temporarily compromise single failure criteria,
administrative controls will ensure the time involved will be limited
to only that required to demonstrate component or system operability.
This LCO provides an acceptable method of restoring equipment to
service for the sole purpose of demonstrating its operability or the
operability of other related equipment. Therefore, this change does not
involve a significant increase in the probability or consequences of
any accident previously evaluated.
2. The proposed change does not create the possibility of a new or
different kind of accident from any previously evaluated.
No modifications were made to the plant due to this amendment
request. The proposed change does not alter the functional
characteristics of any plant component and does not allow any new modes
of operation for any component. This proposed amendment would
facilitate the testing of equipment in its design configuration to
demonstrate operability. The use of TS 3.0.6 would be limited to the
time absolutely necessary to perform the test. Therefore, this change
does not create the possibility of a new or different kind of accident
from any previously evaluated.
3. The proposed change does not involve a significant reduction in
a margin of safety.
The addition of TS 3.0.6 is considered necessary to establish an
allowance that is not formally recognized in the current TSs. Without
this allowance, situations can arise in which certain components could
not be restored to operable status without requiring a plant shutdown.
It is not the intent that the TSs preclude the return to service of a
component to confirm its operability. This allowance is deemed to
represent a more stable, safe operation than requiring a plant shutdown
to complete the restoration and confirmatory testing. The time period
during which the equipment is returned to service in conflict with the
requirements of the TS Action statement is limited to the time
absolutely necessary to perform the indicated surveillance requirement.
TS 3.0.6 does not provide time to perform any other preventive or
corrective maintenance. The period of time during which the equipment
is returned to service will be limited by administrative controls and
is considered very small. Therefore, the probability of an accident
during that time period is also very small and is considered to be
insignificant. Thus, it can be concluded that the proposed change does
not affect the current margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: March 9, 1999.
Description of amendment request: The proposed change would modify
the Technical Specifications to increase the inservice inspection
interval, and reduce the scope of volumetric and surface examinations
for the reactor coolant pump flywheels.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
The Nuclear Regulatory Commission has provided standards in 10
CFR 50.92(c) for
[[Page 24197]]
determining whether a significant hazard exists due to a proposed
amendment to an Operating License for a facility. A proposed
amendment involves no significant hazards consideration if operation
of the facility in accordance with the proposed changes would: (1)
Not involve a significant increase in the probability or
consequences of an accident previously evaluated; (2) Not create the
possibility of a new or different kind of accident from any accident
previously evaluated; or (3) Not involve a significant reduction in
a margin of safety. The Davis-Besse Nuclear Power Station has
reviewed the proposed changes and determined that a significant
hazards consideration does not exist because operation of the Davis-
Besse Nuclear Power Station, (DBNPS) Unit No. 1, in accordance with
these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no accident initiators,
conditions, or assumptions are affected by the proposed changes to
Technical Specification Surveillance Requirement 4.4.10.1.a in the
frequency and scope of volumetric and surface examinations for the
Reactor Coolant Pump (RCP) motor flywheels.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because changes in the frequency and
scope of volumetric and surface examinations for the RCP motor
flywheels will not affect any previously evaluated accidents.
Accidents associated with failure of the flywheel were not evaluated
in the DBNPS Updated Safety Analysis Report (USAR). The design,
fabrication, and testing of flywheels in accordance with the
guidance found in NRC Regulatory Guide 1.14, ``Reactor Coolant Pump
Flywheel Integrity,'' Revision 1, August 1975, minimizes the
potential for flywheel failure. The proposed changes have been
demonstrated to maintain conservative testing requirements for the
flywheels.
2. Not create the possibility of a new or different kind of
accident from any previously evaluated because changes in the
frequency and scope of volumetric and surface examinations for the
RCP motor flywheels will not affect the reliability of RCP motor
flywheels. No new failure mode is introduced since the proposed
changes do not involve a modification or change in operation of any
plant systems, structures, or components.
3. Not involve a significant reduction in the margin of safety.
As shown in Westinghouse Topical Report WCAP-14535A, ``Topical
Report on Reactor Coolant Pump Flywheel Inspection Elimination,''
November 1996, RCP motor flywheels have been inspected for twenty
years without any service induced flaws being identified.
Additionally, the analyses demonstrated that the flywheels are
manufactured from high quality steel, have a high fracture
toughness, and have a very high flaw tolerance. The topical report
indicates that the flywheels could be operated for forty years
without inspection, and there would be no significant increase in
the probability of failure of the flywheels. However, inspections
are proposed to continue at a frequency of once every ten years as a
conservative measure. Thus, the margin of safety is not reduced
significantly by the proposed change in inspection frequency.
Based on the above, the Davis-Besse Nuclear Power Station has
determined that the License Amendment Request does not involve a
significant hazards consideration. As this License Amendment Request
concerns a proposed change to the Technical Specifications that must
be reviewed by the Nuclear Regulatory Commission, this License
Amendment Request does not constitute an unreviewed safety question.
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.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, OH 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Anthony J. Mendiola.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: March 31, 1999.
Description of amendment request: The proposed change would modify
Cooper Nuclear Station's technical specification administrative
controls for unit staff qualifications for the shift supervisor, senior
operator, licensed operator, shift technical advisor, and radiological
manager.
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:
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed change provides enhancement to the current
requirements and clarifies the qualifications and training
requirements for the shift supervisor, senior operator, licensed
operator, shift technical advisor, and Radiological Manager. This
provides additional assurance that these personnel are properly
trained and qualified for their positions; therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change will not create the possibility of a new or
different kind of accident than evaluated in the Updated Safety
Analysis Report (USAR). The proposed change provides enhancement to
the current requirements and clarifies the qualifications and
training requirements for the shift supervisor, senior operator,
licensed operator, shift technical advisor, and Radiological
Manager. The revised administrative controls for unit staff
qualifications are an enhancement to the current requirements;
therefore, the proposed change does not create the possibility of a
new or different kind of accident.
The proposed change will not create a significant reduction in
the margin of safety. The proposed change provides enhancement to
the current requirements and clarifies the qualifications and
training requirements for the shift supervisor, senior operator,
licensed operator, shift technical advisor, and Radiological
Manager. This provides additional assurance that these personnel are
properly trained and qualified for their positions; therefore, the
proposed change will not create 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.
Local Public Document Room location: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Robert A. Gramm.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: March 31, 1999.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Table 3.6.1.2-1, ``Allowable Leak
Rates through Valves in Potential Bypass Leakage Paths,'' by adding two
relief valves, with associated leak rate criteria, to be installed on
the drywell equipment drain line and drywell floor drain line during
the refueling outage in the spring of 2000. Specifically:
(i) For the drywell equipment drain line, the reference to the
inboard isolation valve (2DER*MOV119) would be replaced with a
reference to the isolation valve and its associated relief valve
(2DER*MOV119 and 2DER*RV344);
(ii) For the drywell floor drain line, the reference to the inboard
isolation valve (2DFR*MOV121) would be
[[Page 24198]]
replaced with a reference to the isolation valve and its associated
relief valve (2DFR*MOV121 and 2DFR*RV228); and
(iii) A footnote for both above changes would be added to state,
``For valves 2DER*MOV 119 and 2DER*RV344, and likewise for valves
2DFR*MOV121 and 2DFR*RV228, this limit shall be the combined allowable
leak rate and not the per valve allowable leak rate.''
The two relief valves would be installed to protect the drain line
penetrations against overpressure, consistent with Generic Letter 96-
06, ``Assurance of Equipment Operability and Containment Integrity
During Design-Basis Accident Conditions.'' The allowable leak rates
currently specified in TS Table 3.6.1.2-1 for the drywell equipment and
drywell floor drain line penetrations will not be increased as a result
of the hardware modifications or proposed TS amendment.
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:
The operation of Nine Mile Point Unit 2 [NMP2], in accordance
with the proposed amendment, will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendment will add one relief valve on the drywell
equipment drain line (penetration 2DER*Z40) and one relief valve on
the drywell floor drain line (penetration 2DFR*Z39). These valves
will be installed on piping between the inboard containment
isolation valve and the primary containment wall. These drain lines
represent potential bypass leakage paths from the primary
containment to the environment and are subject to maximum allowable
isolation valve leak rates, as specified in Table 3.6.1.2-1 of the
Technical Specifications (TS). The purpose of adding relief valves
is to protect the piping between the inboard and outboard isolation
valves against thermally induced overpressure under postulated
accident conditions when both isolation valves close, and the fluid
trapped between them may heat up and expand. The new relief valves
and piping will not cause any existing plant design, operating, or
testing limits to be exceeded. The relief valve installations will
meet standards and specifications currently applicable to the
penetrations being modified. The relief valve configuration, set
pressure, and testing meet applicable NRC guidance. No different
precursors or new accident initiators are introduced as the result
of the proposed modification. Therefore, this proposed amendment
does not involve a significant increase in the probability of an
accident previously evaluated.
The existing requirements relating to allowable bypass leakage
for the two penetrations affected by this modification, will not be
changed. No new bypass leakage paths to the environment will be
created and no new failure modes will be introduced. Should the
relief valves open and fail to close, the effectiveness of the
containment and other fission product barriers will not be
compromised. As a result, accident dose rates will remain unchanged
and within the limits of 10 CFR 50, Appendix A, General Design
Criterion 19, and 10 CFR 100. None of the accident assumptions
described in Section 6.2, titled ``Containment Systems'' and Chapter
15, titled ``Accident Analysis,'' of the NMP2 Updated Safety
Analysis Report (USAR) is adversely affected by the proposed
modifications. Therefore, this proposed amendment does not involve a
significant increase in the consequences of an accident previously
evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The isolation valves associated with penetrations 2DER*Z40
(drywell equipment drain line) and 2DFR*Z39 (drywell floor drain
line) perform an accident mitigation function by isolating the
containment during and after certain postulated accidents. The
addition of relief valves between the inboard and outboard isolation
valves will enhance the capability of the existing isolation valves
to perform their function without the risk of failure due to piping
overpressurization. Consistent with the guidance in Generic Letter
96-06, the consequences of a stuck-open relief valve malfunction
have been evaluated and are acceptable. Should the relief valve fail
to close after opening, the existing outboard isolation valve will
perform its function to isolate the containment. Therefore,
operation of NMP2 in accordance with this proposed amendment will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The proposed installation of the relief valves will not
adversely affect primary containment integrity, the maximum
allowable leak rates for the affected penetrations, any other
fission product barriers, or any plant safety/operational limits.
The relief valves will assure that the associated isolation valves
do not fail as the result of piping overpressure during and after
postulated accidents, which will preserve the radiological margin of
safety. Therefore, operation of NMP2 in accordance with the proposed
amendment will not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: S. Singh Bajwa
Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: March 2, 1999.
Description of amendment request: The proposed amendment would
require two service water (SW) pumps and their associated strainers to
be operable to declare a service water system (SWS) loop operable. The
proposed amendment would also (1) modify the existing action statement
to take into account one or more service water pump(s) or strainers
being inoperable and (2) make changes to the appropriate Bases section.
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:
NNECO has reviewed the proposed revision in accordance with 10
CFR50.92 and has concluded that the revision does not involve any
Significant Hazards Considerations (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
satisfied. The proposed Technical Specification revision does not
involve an SHC because the revision would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed TS [Technical Specification] change adds an
additional AOT [allowed outage time] for one of four of the service
water pumps/strainers in the SWS. The capabilities of the SWS were
evaluated in order to ensure that a significant increase in the
probability or consequences of the following previously evaluated
accidents, LOP [loss of power], LOCA [loss-of-coolant accident] with
concurrent LOP and secondary side piping break inside containment,
are precluded by SWS mitigative functions. As the above DBA's
[design basis accidents] are not caused by the failure of the SWS to
operate, the SWS can not affect the probability of these accidents
to occur.
Since both pumps/strainers in each loop are covered by the
ACTION statement in the TS when inoperable (due to failure or
maintenance), and the proposed ACTION statement for two inoperable
service water pumps in a single loop is consistent with the
[[Page 24199]]
current ACTION statement, there is no impact on the capability to
maintain core decay heat removal following a DBA. Further, the
revised TS will improve availability of the SWS. The LCO [limiting
condition for operation] and ACTION statements help ensure that the
SWS, including pumps/strainers, are kept in a condition which allows
it to perform all its design functions including providing core
decay heat removal and the SFP [spent fuel pool] cooling. As such,
there is no affect on the consequences of previously evaluated
accidents.
Thus, it is concluded that the proposed revision does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The SWS is used to remove heat from the reactor plant auxiliary
systems and other systems. Only one of four pumps is required to be
operating during normal plant conditions. In addition, only one 100%
capacity pump is required to provide the necessary flow to mitigate
the consequences of a DBA. This change continues to require two
pumps/strainers per loop to be operable and imposes strict controls
on the AOT for the SWS pumps/strainers via the imposition of the LCO
controls on the SWS. This assures that four service water pumps/
strainers will always be available or the plant will be in an ACTION
STATEMENT. The SWS is used to mitigate the consequences of an
accident and will not cause an accident.
Thus, this proposed revision does 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.
This change will have no impact on the performance of any safety
related system covered by the TS. This change explicitly defines the
number of pumps/strainers required for the SWS to be considered
OPERABLE and the ACTION required which specifies the AOT for
inoperable components. The required flow rate for accident
mitigation continues to be available to all ECCS [emergency core
cooling system] components and their support systems. As such, this
change does not increase the peak clad temperature for a DBA-LOCA.
The proposed Technical Specification change adds an additional
AOT for one of four of the service water pumps/strainers in the SWS.
Two service water pumps/strainers are required to perform the design
function of the SWS; one pump to mitigate the DBA and the other to
reduce the potential of the SFP boiling which could occur if a
service water pump is unavailable for SFP cooling after a design
basis LOCA.
The existing TS Bases states that ``The OPERABILITY of the
Service Water System ensures that sufficient cooling capacity is
available for continued operation of safety-related equipment during
normal and accident conditions. The redundant cooling capacity of
this system, assuming a single failure, is consistent with the
assumptions used in the safety analyses.''
Since this change continues to control the availability of the
SW pumps by placing the system in an ACTION statement with one loop
out of service, then the change will continue to comply with the
existing BASES requirements. Thus it is concluded that the proposed
revision does not involve a significant reduction in the margin of
safety.
In conclusion, based on the information provided, it is
determined that the proposed revision does not involve a SHC.
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.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Section Chief: James W. Clifford.
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Dockets
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: December 24, 1998.
Description of amendment request: Revises the setpoints and limits
of allowable values for loss of power (LOP) instrumentation for 4kV
emergency busses.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The LOP instrumentation provides safety-related electrical
equipment protection. No new equipment is added to the plant as a
result of the proposed changes. Separation of the 4kV emergency
buses from the grid is the only potential transient that previously
existed based on operation of these relays. Based on the revised
Voltage Regulation Study, which incorporates the effects of system
improvements and additional conservatisms, there is no significant
increase in the probability of this separation. The relay time delay
settings are such that the relays will detect and respond to an
actual sustained degradation of voltage, but will not actuate in
response to normal operational voltage fluctuations. No accident
initiators will be impacted by the proposed setpoint changes. All
safety systems will be able to perform their safety functions.
Accident mitigation is achieved by these relays by ensuring adequate
voltage is maintained throughout the Class 1E electrical
distribution system.
The existing allowable values and the proposed allowable values
for Functions 2, 3, 4, and 5 have been analyzed and both values are
acceptable for operation. During implementation of modification 96-
01511 (changing of the relay setpoints), the 4kV buses could be in
one of the three configurations: (a) Both sources have relays set at
the existing setpoints, (b) one set of source relays with the
existing old setpoints and the other set with the proposed revised
setpoints, or (c) both sources have relays set at the proposed
revised setpoints. Each of these configurations is acceptable
because the existing and proposed values satisfy the design limits
established within the setpoint calculation and the Voltage
Regulation Study.
For Function[s] 4 and 5, the present TS has separate entries in
Table 3.3.8.1-1, for the internal and external time delay. This
proposed change will combine these internal and external time delays
for simplicity. The aggregate time delay is the important parameter
and it is the only time delay that is analyzed. The internal time
delay minimizes the relay contact wear and reduces the number of
external time delay relay actuations due to transient voltage dips.
The internal time delay provides no other output functions.
Therefore, there will be no impact on the Class 1E power
distribution system to perform its intended design function.
Therefore, the proposed changes described above, or operation
while modification 96-01511 is being implemented, does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed LOP instrumentation setpoint changes will not
result in any new accidents or operational transients. Separation of
the 4kV emergency buses from the grid is the only potential
transient that previously existed based on operation of these
relays. Based on the revised Voltage Regulation Study, which
incorporates the effects of system improvements and additional
conservatisms, there is no significant increase in the probability
of this separation, and the proposed setpoint changes would not
create the possibility of a new or different kind of accident from
any previously evaluated. The relay time delay settings are such
that the relays will detect and respond to an actual sustained
degradation of voltage, but will not actuate in response to normal
operational voltage fluctuations. The proposed setpoint changes for
these relays and the proposed combining
[[Page 24200]]
of the internal and external time delays will not become initiators
of different types of accidents or transients. Additionally, since
the existing and proposed allowable values for the LOP
instrumentation functions are within the band established by the
Voltage Regulation Study, both values are acceptable for operation
during the implementation of modification 96-01511. Therefore, the
possibility of a new or different kind of accident than previously
evaluated is not created.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
All LOP instrumentation functions will continue to be carried
out. The proposed setpoint and allowable value changes have been
evaluated within the Voltage Regulation Study and the Plant
Electrical Load Study. The relay setpoints have been established
using IISCP setpoint methodology. The setpoint determination
accounts for relay accuracy, potential transformer accuracy,
measurement and test equipment accuracy, and margin above the design
limit established within the Voltage Regulation Study. The proposed
setpoint changes for these relays and the proposed combining of the
internal and external time delays will not involve a significant
reduction in a margin of safety. Additionally, since the existing
and proposed allowable values for the LOP instrumentation functions
are within the band established by the Voltage Regulation Study,
both values are acceptable for operation during the implementation
of modification 96-01511. Therefore, having both values during the
implementation of modification 96-01511 does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Attorney for Licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Project Director: Elinor G. Adensam
PECO Energy Company, Public Service Electric and Gas Company, Delmarva
Power and Light Company, and Atlantic City Electric Company, Dockets
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2
and 3, York County, Pennsylvania
Date of application for amendments: February 12, 1999.
Description of amendment request: Administrative changes to correct
typographic errors in Technical Specifications (TS) introduced in
previous amendments.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes correct typographical errors and are
administrative only and do not impact the operation of the facility.
In each case, the action of the intended TS requirements were
satisfactorily completed when the change was implemented. These
corrections are administrative only and have no effect on any
previously evaluated accident scenario. The changes will not alter
the operation of equipment assumed to be available for the
mitigation of accidents or transients, nor will they alter the
operation of equipment important to safety previously evaluated.
Therefore, the changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes correct typographical errors and are
administrative only and will not involve any physical changes to the
plant SSCs [systems, structures, or components]. In each case, the
action of the intended TS requirements were satisfactorily completed
when the change was implemented. These corrections are
administrative only and have no effect on any previously evaluated
accident scenario. The proposed changes do not allow operation in
any mode that is not already evaluated. The changes will not alter
the operation of equipment important to safety previously evaluated.
Therefore, the changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes correct typographical errors and are
administrative only and will not affect the manner in which the
facility is operated, or change equipment or features which affect
the operational characteristics of the facility. The proposed
changes have no impact on any safety analysis assumptions or margins
of safety.
Therefore, these proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
PA 17105.
Attorney for Licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,
PA 19101.
NRC Section Chief: James W. Clifford.
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: January 21, 1999.
Description of amendment request: The proposed amendments would
change Technical Specification Tables 3.3.6.1-1 and 3.3.6.2-1 by
increasing the Allowable Values for the high radiation trip for the
exhaust monitors for the reactor building and the refueling. The
January 21, 1999, amendment request supercedes the July 22, 1998,
amendment request which was noticed in the Federal Register on August
26, 1998 (63 FR 45529).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1). Do the proposed changes involve a significant increase in
the probability or consequences of an accident previously evaluated?
The Unit 1 and Unit 2 reactor building and refueling floor
ventilation exhaust radiation monitors perform no function in
preventing, or decreasing the probability of, a previously evaluated
accident. The monitors are designed to monitor ventilation exhaust
for indications of a release of radioactive material resulting from
a design basis accident and initiate appropriate protective actions.
Because the proposed changes affect only the ventilation exhaust
radiation monitors, the probability of an accident previously
evaluated remains the same.
The function of the reactor building and the refueling floor
ventilation exhaust radiation monitors, in combination with other
accident mitigation systems, is to limit fission product release
during and following postulated design basis accidents. The proposed
new Allowable Values for the high radiation trip will continue to
ensure the offsite doses resulting from a design basis accident do
not exceed the NRC-approved
[[Page 24201]]
licensing basis. Therefore, the proposed changes do not involve a
significant increase in the consequences of an accident previously
evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes increase the radiation level at which the
ventilation exhaust monitors actuate; however, the manner in which
their actuation logic functions and the systems that isolate or
actuate as a result are unaffected by the proposed changes.
Furthermore, the ventilation exhaust monitors will continue to
perform their design function of limiting offsite doses to NRC-
approved licensing limits at the higher Allowable Values. Therefore,
the proposed changes cannot create the possibility of a new or
different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
The Bases for Unit 1 and Unit 2 Technical Specifications Tables
3.3.6.1-1 and 3.3.6.2-1 state that the Allowable Values for the
reactor building and refueling floor ventilation exhaust radiation
monitors ``are chosen to ensure radioactive releases do not exceed
offsite dose limits.'' The proposed Allowable Values ensure the
radiation monitors actuate at a radiation level sufficient to ensure
offsite doses are within the NRC-approved licensing basis. The
proposed Allowable Values comply with the margin of safety defined
in the Technical Specifications Bases for the ventilation exhaust
radiation monitors; therefore, the proposed changes do not reduce 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.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC.
NRC Section Chief: Richard L. Emch, Jr.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 30, 1999.
Description of amendment request: The licensee has proposed to
relocate Technical Specification 3/4.3.3.4, ``Meteorological
Instrumentation,'' and its associated Bases to the Technical
Requirements Manual (TRM). Because the TRM is incorporated within the
South Texas Project updated final safety analysis report (UFSAR) for
the units, changes to the requirements on the meteorological
instrumentation that would be relocated to the TRM would be controlled
in accordance with 10 CFR 50.59.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The affected system and components [i.e., meteorological
monitoring instrumentation] are not assumed as initiators of
analyzed events, and are not assumed to mitigate accident or
transient events. The requirements and surveillances for [this
affected system] and components will be relocated from the Technical
Specifications to the Technical Requirements Manual, which is
incorporated in the South Texas Project UFSAR and will be maintained
pursuant to 10 CFR 50.59. In addition, the Meteorological Monitoring
System components are addressed in existing surveillance procedures
which are also controlled by 10 CFR 50.59 and subject to the change
control provisions imposed by plant administrative procedures, which
endorse applicable regulations and standards. The associated changes
to the Technical Specification Index are administrative. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This change does not involve a physical alteration of the plant
(no new or different type of equipment will be installed) or make
changes in the methods governing normal plant operation. This change
will not impose different requirements, and adequate control of
information will be maintained. Furthermore, this change will not
alter assumptions stated in the safety analysis or licensing basis.
The associated changes to the Technical Specification Index are
administrative. Therefore, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
This change will not reduce a margin of safety because the
change has no impact on any safety analysis assumptions. In
addition, the relocated requirements and surveillances for the
affected structures, systems, and components remain the same as the
existing Technical Specifications. Because any future changes to
these requirements or the surveillance procedures will be evaluated
per the requirements of 10 CFR 50.59; there is no [significant]
reduction in a margin of safety. The associated changes to the
Technical Specification Index are administrative and have no
potential effect on the margin of safety. Therefore, this change
does not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas
77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
NRC Section Chief: Robert A. Gramm.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: March 2, 1999 (TS 98-05).
Brief description of amendments: The proposed amendments would
change the SQN Operating Licenses DPR-77 (Unit 1) and DPR-79 (Unit 2)
by eliminating a requirement to have an Independent Safety Engineering
Group (ISEG), conditions imposed by NUREG-0737. Because of evolution
through numerous reorganizations and reassignments, these license
conditions are no longer necessary and the Tennessee Valley Authority
(TVA, the licensee) proposes deleting them.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), 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 possibility of occurrence or the consequences for an
accident or malfunction of equipment is not increased. The ISEG
function is one of ``oversight'' only.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
A possibility for an accident or malfunction of a different type
than any evaluated previously in SQN's Final Safety Analysis Report
is not created by the proposed elimination of the ISEG; nor is the
possibility for an accident or malfunction of a different type. The
ISEG function is one of ``oversight'' only.
[[Page 24202]]
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed amendment will not involve a significant reduction
in the margin of safety. The ISEG function is one of ``oversight''
only.
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.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
NRC Section Chief: Sheri R. Peterson.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: February 12, 1999 (TXX-99022).
Brief description of amendments: The proposed changes would modify
the steam generator tube inspection requirements and acceptance
criteria to implement the 1.0-volt repair criteria for steam generator
tubes affected by outer diameter stress corrosion cracking (ODSCC)
according to Nuclear Regulatory Commission (NRC) Generic Letter 95-05
(``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes
Affected by Outside Diameter Stress Corrosion Cracking'') at Comanche
Peak Unit 1. Also proposed is the use of a voltage-dependent
probability of detection; the methodology was originally submitted to
the NRC by the Nuclear Energy Institute in 1996.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of Comanche Peak Unit 1 in accordance with the
proposed license amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Tube burst criteria are inherently satisfied during normal
operating conditions due to the proximity of the tube support plate
[TSP]. Test data indicates that tube burst cannot occur within the
TSP, even for tubes which have 100% through-wall electric discharge
machining notches, 0.75 inch long, provided that the TSP is adjacent
to the notched area. Since tube to tube support plate proximity
precludes tube burst during normal operating conditions, use of the
criteria must retain tube integrity characteristics which maintain a
margin of safety of 1.43 times the bounding faulted condition (Steam
Line Break) pressure differential. As previously stated, the RG
[Regulatory Guide] 1.121 criterion requiring maintenance of a safety
factor of 1.43 times the Steam Line Break pressure differential on
tube burst is satisfied by \3/4\'' diameter tubing with bobbin coil
indications with signal amplitudes less than 4.7 volts, regardless
of the indicated depth measurement. At the FDB [flow distribution
baffle], a safety factor of 3 against the normal operating condition
at power is applied. Here a voltage of 3.34 volts satisfies the
burst capability recommendation.
The upper voltage repair limit (VURL) will be
determined prior to each outage using the most recently approved NRC
database to determine the tube structural limit (VSL).
The structural limit is reduced by allowances for nondestructive
examination (NDE) uncertainty (VNDE) and growth
(VGr) to establish VURL. As an example, the
NDE uncertainty component of 20% and a voltage growth allowance of
30% per full power year can be utilized to establish a
VURL of 3.13 volts for TSP indications, 2.22 volts for
the FDB indications. The 20% NDE uncertainty represents a
squareroot-sum-of-the-squares (SRSS) combination of probe wear
uncertainty and analyst variability.
The flaw growth allowance should be an average growth rate or
30% per effective full power year, whichever is larger. The 30%
growth allowance used to determine VURL is conservative
for the current conditions at Comanche Peak Unit 1. The average
growth of the bobbin indication voltages observed at the last
inspection is determined to be 0.14 volts, or 24.6% voltage growth.
This value is a conservative representation of the growth trends at
Comanche Peak Unit 1 as not all steam generators were inspected at
end of cycle 3 and end of cycle 4, and the largest reported voltage
growths represent more than one cycle of actual plant operation. The
most current NRC approved database, contained in EPRI [Electric
Power Research Institute] NP-7480-L, Addendum 1, was used to
establish the VURL values for the FDB and TSP
intersections. Once approved by the NRC, the industry protocol for
updating the database will be followed by TU Electric, ensuring that
the most current database is utilized for all future applications of
the criteria.
Also, assuming the criteria was applied at the last inspection
at Comanche Peak Unit 1, using conservative growth projections as
described in Reference 2 [of the February 12, 1999, application],
the conditional burst probability at end of cycle 6 is determined to
be 1.7 x 10-4, which is well within the GL 95-05
reporting limit of 1 x 10-2.
Relative to the expected leakage during accident condition
loadings, it has been previously established that a postulated main
Steam Line Break outside of containment but upstream of the MSIV
[main steam isolation valve] represents the most limiting
radiological condition relative to the plugging criteria. In support
of implementation of the revised plugging limit, it will be
determined whether the distribution of cracking indications at the
tube support plate intersections during future cycles are projected
to be such that primary to secondary leakage would result in site
boundary doses within 10CFR100 guidelines and control room doses
within the GDC [General Design Criterion]-19 limit. A separate
calculation has determined this allowable Steam Line Break leakage
limit to be 27.79 gpm in the faulted loop assuming a RCS [reactor
coolant system] dose equivalent I-131 concentration of 1.0 microCi/
gm. The establishment of the 27.79 gpm leak rate value is controlled
by the 0 to 2 hour offsite dose at the site boundary for the
accident initiated iodine spike case, not the control room dose. For
this case, the site boundary thyroid dose approaches, but is bounded
by, the 30 Rem limit recommended in NUREG-0800 [``Standard Review
Plan''].
The methods for calculating the radiological dose consequences
are also revised for this application. Rather than basing the
calculated thyroid dose consequences on conversion factors from TID-
14844, [``Calculation of Distance Factors for Power and Test Reactor
Sites''] factors obtained from ICRP-30 [International Commission on
Radiation Protection Publication 30] are used. The use of ICRP-30
dose conversion factors in this application has been previously
accepted by the NRC. Although the use of ICRP-30, relative to the
TID-14844, results in lower calculated thyroid doses for this
application, the NRC has previously determined that the ICRP-30
factors retain adequate conservatism.
In summary, due to the methodology used to determine the maximum
allowable, accident-initiated leak rate (prescribed in Section 2.b.4
of Generic Letter 95-05), the calculated radiological consequences
at the EAB [exclusion area boundary] and LPZ [low population zone]
are larger than previously reported for the postulated steamline
break event. However, the calculated radiological consequences
remain in compliance with NUREG-0800 and GDC-19. Therefore, it is
concluded that the proposed changes do not result in a significant
increase in the radiological consequences of an accident previously
analyzed.
The removal from the FSAR [final safety analysis report] of the
steamline break radiological dose consequences calculation typically
identified as a ``5% failed fuel'' scenario does not affect the
probability or consequences of any accident previously considered.
For CPSES [Comanche Peak Steam Electric Station], no accident-
induced fuel failures are predicted; therefore, consistent with
NUREG-0800, this scenario is not required to be analyzed or
presented in the FSAR.
In summary, because the implementation of the 1.0 volt voltage-
based plugging criteria at Comanche Peak Unit 1 does not adversely
affect steam generator tube integrity and implementation will be
shown to result in acceptable radiological dose consequences, the
proposed Technical Specification change does not result in any
increase in the probability or consequences of an accident
previously evaluated within the Comanche Peak FSAR.
(2) The proposed license amendment does not create the
possibility of a new or different
[[Page 24203]]
kind of accident from any accident previously evaluated.
Implementation of the proposed steam generator tube 1.0 volt
plugging limit does not introduce any significant changes to the
plant design basis. Neither a single or multiple tube rupture event
would be expected in a steam generator in which the plugging limit
has been applied (during all plant conditions).
The bobbin probe voltage-based tube plugging criteria of 1.0
volt is supplemented by: enhanced eddy current inspection guidelines
to provide consistency in voltage normalization, a 100% eddy current
inspection sample size at the tube support plate elevations, and RPC
[rotating pancake coil] inspection requirements for the larger
indications left in service to characterize the principal
degradation as ODSCC. TU Electric will implement a maximum normal
operating condition primary to secondary leakage rate limit of 150
gpd (0.1 gpm--at room temperature) per steam generator to help
preclude the potential for excessive leakage during all plant
conditions. The 150 gpd leakage limit is more restrictive than the
standard operating leakage limit (of 500 gpd) and is intended to
provide additional margin to accommodate a stress corrosion crack
which might grow at a greater than expected rate or unexpectedly
extend outside the thickness of the tube support plate. Leakage
trending capability consistent with EPRI Report TR-04788, ``PWR
Primary-to-Secondary Leak Guidelines'', has been implemented at
Comanche Peak Unit 1.
As steam generator tube integrity upon implementation of the 1.0
volt plugging limit continues to be maintained through in-service
inspection and primary to secondary leakage monitoring, the
possibility of a new or different kind of accident from any accident
previously evaluated is not created.
(3) The proposed license amendment does not involve a
significant reduction in margin of safety.
The use of the voltage-based bobbin probe tube support plate
elevation plugging criteria at Comanche Peak Unit 1 maintains steam
generator tube integrity commensurate with the criteria of
Regulatory Guide 1.121. Regulatory Guide 1.121 describes a method
acceptable to the NRC staff for meeting GDCs 14, 15, 31, and 32 by
reducing the probability or the consequences of steam generator tube
rupture. This is accomplished by determining the limiting conditions
of degradation of steam generator tubing, as established by
inservice inspection, for which tubes with unacceptable cracking
should be removed from service. Upon implementation of the proposed
criteria, even under the worst case conditions, the occurrence of
ODSCC at the tube support plate elevations is not expected to lead
to a steam generator tube rupture event during normal or faulted
plant conditions. The end of cycle distribution of crack indications
at the tube support plate elevations is confirmed to result in
acceptable primary to secondary leakage during all plant conditions
and that radiological consequences are not adversely impacted.
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.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019 Attorney for licensee: George L. Edgar,
Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC
20036.
NRC Section Chief: Robert A. Gramm
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.
Date of amendment request: February 16, 1999.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) Sections 3.6, 3.9, and 3.16
and the associated Bases for those sections for Units 1 and 2. The
proposed changes would consolidate the auxiliary feedwater (AFW) cross-
connect requirements by relocating the electrical power requirements
from Section 3.16 to Section 3.6. The proposal also would clarify the
TS with regard to permitting simultaneous entry into certain conditions
of operation on Units 1 and 2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1--Operation of Surry Units 1 and 2 in accordance with
the proposed TS change does not involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated.
The proposed TS change is administrative in nature, and station
operations are not being affected. The accidents considered relative
to this proposed TS change are Rupture of Main Steam Pipe, Loss of
All AC Power, and Loss of Feedwater. The probability of occurrence
of these accidents has been previously evaluated to support Surry TS
Amendment 143/140. The NRC reviewed the PSA [probabilistic safety
analysis] basis during issuance of TS Amendment 143/140 and found it
acceptable. The probability of occurrence of these accidents has
been recently reviewed relative to this proposed TS change. It has
been concluded that the proposed TS change is consistent with the
existing analyses and evaluations and, therefore, will not increase
the probability of occurrence of the identified accidents.
The consequences of the accidents identified above were also
previously evaluated to support Surry TS Amendment 143/140. The PSA
considerations included the AFW cross-connect capability, diesel
generator dependencies, various LCO [limiting condition for
operation] time periods, and a HELB [high energy line break] in the
vicinity of the AFW Pumps. The previous evaluation was recently
reviewed relative to this proposed TS change. This review determined
that the proposed TS change is consistent with the design and
licensing bases supporting the existing Technical Specifications.
The proposed TS change is also consistent with the existing analyses
and evaluations, the consequences of which bound any potential
consequences of the proposed TS change. Therefore, the proposed TS
change will not increase the consequences of the identified
accidents.
Criterion 2--The proposed TS change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The possibility for a new or different type of accident than any
previously evaluated is not created since the considerations in the
PSA and evaluations performed to support TS Amendment 143/140 are
not changed by the proposed administrative TS change. The proposed
TS change is consistent with the design and licensing bases
supporting the existing Technical Specifications. Furthermore,
station operations and plant equipment are not being affected and,
therefore, the proposed TS change does not create any new failure
modes or accident precursors.
Criterion 3--The proposed TS change does not involve a
significant reduction in a margin of safety.
The proposed administrative change to Surry Technical
Specifications clarifies the requirements (limiting conditions for
operation (LCO) and action statements) relating to the Auxiliary
Feedwater (AFW) cross-connect by relocating the emergency power
source requirements of TSs 3.16.A.8 and 3.16.B.4 to TS 3.6. The
proposed TS change does not alter the current TS requirements or
bases, as well as maintains the Surry licensing and design basis.
The proposed change does not affect either station operations or
plant equipment, hence the availability of equipment for the
mitigation of accidents is not decreased. Furthermore, the
assumptions governing the accident analyses remain unchanged, and
the consequences of the existing analyses and evaluations remain
bounding. This is an administrative change and as such does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
[[Page 24204]]
Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams,
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia
23219.
NRC Section Chief: Richard L. Emch, Jr.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: February 16, 1999.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) Section 4.2 for Units 1 and 2
to relax the surveillance requirements for reactor coolant pump (RCP)
flywheels. The flywheels provide extended reactor coolant flow
coastdown capability if electric power for the RCPs is lost. Currently,
the flywheels are subjected to an inspection program that meets the
requirements of NRC Regulatory Guide 1.14, Revision 1, dated August
1975. The inspections include an ultrasonic examination (UT) of areas
of high stress concentration at the bore and keyway every three years,
and complete UT every 10 years. The proposed change would require only
a 10-year UT, based upon an analysis presented in a Westinghouse
topical report (WCAP-14535A) which has been reviewed and accepted by
NRC staff.
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 reduction of the inspection requirements for the reactor
coolant pump flywheels, as generically approved by the NRC and
technically supported by WCAP-14535A, does not significantly
increase the probability of an accident previously evaluated in the
safety analysis report. The results of WCAP-14535A have been
reviewed and evaluated with the technical basis accepted for
referencing in license applications by the NRC in their letter
entitled ``Acceptance for referencing of Topical Report WCAP-14535,
Topical Report on Reactor Coolant Pump Flywheel Inspection
Elimination,'' dated September 12, 1996.
The proposed Technical Specification change reduces the
surveillance requirements (inspection) on the RCP flywheel. There is
no change in the method of plant operation or system design. The
WCAP-14535A report establishes that the proposed change has a
negligible affect on the probability that the flywheel will fail
given that the flywheels received preservice and inservice
examinations as required previously. Therefore, the proposed change
does not increase the probability of occurrence or consequences of
any previously analyzed accident.
b. The proposed change to reduce the inspection requirements for
the RCP flywheels as generically approved by the NRC and supported
by WCAP-14535A does not create the possibility of a new or different
kind of accident from any accident previously evaluated in the
safety analysis report.
The proposed surveillance requirements (inspection) only reduce
the inspection requirements/frequency for the reactor coolant pump
flywheels, and there is no change in the method of plant operation
or system design.
c. The proposed change reducing the inspection of the RCP
flywheels as generically approved by the NRC and supported by WCAP-
14535A, does not impact the accident analysis assumptions or the
basis of any Technical Specification. As previously stated, the
analysis performed in the WCAP-14535A report established that the
affect on flywheel failure probability was negligible given that the
initial preservice and inservice inspections under the current
requirements were performed. Therefore, the proposed change in
surveillance (inspection) frequency does not involve a significant
reduction in the margin of safety.
The analysis provided herein demonstrates that the proposed
amendment to the Surry Technical Specifications does not involve a
significant increase in the probability or consequences of a
previously evaluated accident, does not create the possibility of a
new or different kind of accident, and does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams,
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia
23219.
NRC Section Chief: Richard L. Emch, Jr.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of amendment request: April 12, 1999 (TSCR 212).
Description of amendment request: The purpose of the proposed
amendments is to update references in the Technical Specifications. The
update is necessary to reflect relocation of referenced information in
the Final Safety Analysis Report.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed amendment corrects references within the Technical
Specification requirements such that they refer to the correct
information in the updated Final Safety Analysis Report (FSAR). The
references changed due to relocation of the information within the
FSAR. The Technical Specification requirements and intent are not
changed. Therefore, these changes are administrative only and do not
change the design or operation of the Point Beach Nuclear Plant
[PBNP]. Operation of PBNP in accordance with the proposed amendments
cannot increase the probability or consequences of an accident
previously evaluated.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments will not create the possibility of a new or
different kind of accident previously evaluated.
The proposed changes are administrative only and therefore do
not materially change any requirements for the design or operation
of PBNP. Therefore, operation in accordance with the proposed
changes cannot create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not create a significant reduction in a
margin of safety.
The proposed changes are administrative only; correcting
references within the Technical Specification requirements. No
requirement on the operation or design of the facility is being
changed. Therefore, there is no 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.
Local Public Document Room location: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: George F. Dick, Jr., Acting.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the
[[Page 24205]]
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, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of application for amendments: November 19, 1998.
Brief description of amendments: The amendments revised Technical
Specification 3.7.6 ``Service Water (SRW) System'' to allow operation
of Calvert Cliffs Unit Nos. 1 and 2 with one SRW plate and frame heat
exchanger in a subsystem secured and removing one containment air
cooler from service to enable the affected SRW subsystem to remain
operable.
Date of issuance: April 14, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 230 and 206.
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 16, 1998 (63
FR 69333). The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated April 14, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendment request: November 1, 1996, as supplemented May
22, 1998, September 14, 1998, January 4, 1999, and March 19, 1999.
Brief description of amendment: The amendment modified the
Technical Specifications for the Brunswick Steam Electric Plant, Units
1 and 2, to extend the Allowed Outage Time for 4.16kV AC balance of
plant buses and the AC electrical power distribution system load group
buses.
Date of issuance: April 15, 1999.
Effective date: April 15, 1999.
Amendment Nos.: 205 and 235.
Facility Operating License Nos. DPR-71 and DPR-62: Amendment
revises the Technical Specifications.
Date of initial notice in Federal Register: February 11, 1998 (63
FR 6977). The supplemental submittals of May 22, 1998, September 14,
1998, January 4, 1999, and March 19, 1999, contained clarifying
information only, and did not change the initial no significant hazards
consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 15, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: October 14, 1998.
Brief description of amendment: The amendment modifies the
acceptance criterion for Surveillance Requirement 3.4.14.2 from the
setpoint value of 465 psig to the analytical limit for the residual
heat removal system of 474 psig reactor coolant system pressure.
Date of issuance: April 20, 1999.
Effective date: April 20, 1999.
Amendment No. 182.
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 4, 1998 (63 FR
59587).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 20, 1999.
No significant hazards consideration comments received: No
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of application for amendment: September 3, 1997, as
supplemented March 13, 1998, and March 18, 1999.
Brief description of amendment: The amendment revises the technical
specifications to delete snubber operability requirements, action
requirements for inoperable snubbers, and snubber testing requirements.
The snubber testing requirements have been relocated to the Palisades
Operating Requirements Manual.
Date of issuance: April 13, 1999.
Effective date: April 13, 1999, and shall be implemented within 60
days.
Amendment No.: 185.
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 8, 1998 (63 FR
17222). The March 18, 1999, submittal requested a 60-day allowance for
implementation of the amendment. This change was within the scope of
the original Federal Register notice and did not change the staff's
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 13, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423-3698.
Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power
Plant, Unit 1, Monroe County, Michigan
Date of amendment request: July 17, 1998 (Reference NRC-98-0044).
Brief description of amendment: This amendment revises the Enrico
Fermi
[[Page 24206]]
Atomic Power Plant, Unit 1, License to allow possession of a nominal
amount of special nuclear material.
Date of issuance: April 15, 1999.
Effective date: On the date of issuance of this amendment and must
be fully implemented no later than 60-calendar days from the date of
issuance.
Amendment No.: 16.
Facility Operating License No. DPR-9: Amendment revised the License
by adding new Part 2.B.4 to the License.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56240). The NRC's related evaluation of the amendment is contained in a
Safety Evaluation dated April 15, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: March 15, 1999, and
supplemented by letter dated March 17, 1999.
Brief description of amendments: The amendments delete from the
joint Technical Specifications Section 3.3.7, ``Control Room Area
Ventilation System (CRAVS) Actuation Instrumentation,'' and Section
3.3.8, ``Auxiliary Building Filtered Ventilation Exhaust System
(ABFVES) Actuation Instrumentation.'' These surveillance requirements
are not applicable to Catawba because the sections do not reflect the
design of the Catawba units.
Date of issuance: April 8, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1--177; Unit 2--169.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in FEDERAL REGISTER: March 24, 1999 (64 FR
14274). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: February 18, 1999.
Brief description of amendments: The amendments revise the
Technical Specifications Surveillance Requirement (SR) 3.6.16.1
regarding surveillance of reactor building access openings, SR 3.6.16.3
regarding surveillance of reactor building structural integrity, and
Administrative Controls 5.5.2 regarding the Containment Leakage Rate
Testing Program. The revised requirements would provide scheduling
flexibility without decreasing quality and safety margin.
Date of issuance: April 9, 1999.
Effective date: As of the date of issuance, to be implemented
within 30 days from the date of issuance.
Amendment Nos.: 178--Unit 1; 170--Unit 2.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 10, 1999 (64 FR
11961). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 9, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley Power
Station, Unit No. 1, Shippingport, Pennsylvania
Date of application for amendment: November 11, 1998, as
supplemented February 26, 1999.
Brief description of amendment: The amendment modified License
Condition 2.C(9) to allow, on a one-time only, extension of the steam
generator inspection interval in Technical Specification Surveillance
4.4.5.3.b. This will allow the steam generator inspection interval to
coincide with the thirteenth refueling outage or the end of 500
effective full power days, whichever occurs sooner.
Date of issuance: April 16, 1999.
Effective date: As of date of issuance, to be implemented within 60
days.
Amendment No: 221.
Facility Operating License No. DPR-66. Amendment revised the
License.
Date of initial notice in Federal Register: December 2, 1998 (63 FR
66593). The February 26, 1999, letter provided additional information
but did not change the initial proposed no significant hazards
consideration determination or expand the amendment request beyond the
scope of the initial notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 16, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: B.F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas
Nuclear One, Units 1 and 2, Pope County, Arkansas
Date of amendment request: June 28, 1996, as supplemented by
letters dated February 23 and March 15, 1999.
Brief description of amendments: The amendments revise the
Technical Specifications to permit the containment equipment hatch to
be open during handling of irradiated fuel in containment and core
alterations provided that the capability for closure is maintained.
Date of issuance: April 16, 1999.
Effective date: As of the date of issuance, and shall be
implemented within 30 days of issuance.
Amendment Nos.: Unit 1--Amendment No. 195; Unit 2--Amendment No.
203.
Facility Operating License Nos. DPR-51 and NPF-6: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 14, 1996 (61 FR
42280). The February 23 and March 15, 1999, letters provided clarifying
information that did not change the scope of the original application
and the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 16, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: April 30, 1998.
Brief description of amendment: The amendment revises the single
largest post-accident load capable of being supplied by the diesel
generators and relocates this value to the Bases for Technical
Specification (TS)
[[Page 24207]]
Surveillance 4.8.1.1.2.c.3. TS Surveillance 4.8.1.1.2.c.3 has been
revised to refer to ``the single largest post-accident load'' rather
than a specific numerical value for diesel generator load reject
testing. This change is consistent with the guidance provided in NUREG-
1432 , ``Improved Standard Technical Specifications for Combustion
Engineering Plants.''
Date of issuance: April 21, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment No.: 204.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56241). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 21, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: November 13, 1997.
Brief description of amendment: The amendment changes the Appendix
A Technical Specifications (TSs) by revising TS 6.8.4.a, Primary
Coolant Sources Outside Containment, to add portions of the containment
vacuum relief and primary sampling systems to the list of systems
included in the Primary Coolant Sources Outside Containment Program.
Date of issuance: April 21, 1999.
Effective date: The license amendment is effective as of its date
of issuance, and shall be implemented within 60 days.
Amendment No.: 150.
Facility Operating License No. NPF-38: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 25, 1998 (63
FR 9601). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 21, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit 3, Citrus County, Florida
Date of application for amendment: October 1, 1997, as supplemented
April 23 and November 17, 1998 and February 19, 1999.
Brief description of amendment: The changes specify criteria for
evaluating the growth of pit-like intergranular attack steam generator
tube degradation identified in tubes in the ``B'' once-through steam
generator (OTSG). Florida Power Corporation also requested to amend the
Improved Technical Specifications to clarify the date by which the OTSG
inservice inspection results are required to be submitted to the NRC.
Date of issuance: April 8, 1999.
Effective date: April 8, 1999.
Amendment No.: 172.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 22, 1997 (62 FR
54873). The supplemental letters dated April 23 and November 17, 1998,
and February 19, 1999 did not change the original no significant
hazards determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit 3, Citrus County, Florida
Date of application for amendment: September 9, 1997, as
supplemented November 7 and 25, 1997, and January 20 and October 30,
1998.
Brief description of amendment: The amendment proposed to revise
the Final Safety Analysis Report (FSAR) analysis of the Makeup System
letdown line failure accident. The revised analysis models the event as
being terminated by manual operator action to isolate the line whereas
the original analysis models an automatic isolation of the break.
Date of issuance: April 13, 1999.
Effective date: April 13, 1999.
Amendment No.: 173.
Facility Operating License No. DPR-72: Amendment approves changes
to the Final Safety Analysis Report.
Date of initial notice in Federal Register: September 24, 1997 (62
FR 50005). The supplemental letters dated November 7 and 25, 1997,
January 20, 1998, and October 30, 1998, did not change the original
proposed no significant hazards consideration determination, or expand
the scope of the amendment request as originally noticed.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 13, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Nuclear Generating Plant, Unit 3, Citrus County, Florida
Date of application for amendment: October 30, 1998, as
supplemented April 7, 1999.
Brief description of amendment: Changes the Crystal River Unit 3
Technical Specifications to delete a note regarding the number of
required channels for the Degrees of Subcooling function, and to
subdivide the Core Exit Temperature (Backup) function into two new
functions in Table 3.3.17-1, Post-Accident Monitoring Instrumentation.
Date of issuance: April 20, 1999.
Effective date: April 20, 1999.
Amendment No.: 174.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 13, 1999 (64 FR
2246). The April 7, 1999, supplement did not affect the original no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 20, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
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: May 27, 1998, as supplemented
October 9, 1998.
Brief description of amendment: Deletes the requirement for
operability of the safety injection tanks in Mode 4 of reactor
operation.
Date of Issuance: April 8, 1999.
Effective Date: Amendment is effective within 30 days of receipt.
Amendment No.: 100.
[[Page 24208]]
Facility Operating License No. NPF-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 29, 1998 (63 FR
40556). The October 9, 1998 supplemental letter provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: November 25, 1998, as
supplemented February 12, 1999.
Brief description of amendment: The amendment approves the proposed
surveillance Technical Specifications related to the once through steam
generator inservice inspections to be completed during the 13R
refueling outage in fall 1999. Related TS Bases changes are also
included.
Date of issuance: April 13, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 209.
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 16, 1998 (63
FR 69342).
The February 12, 1999, submittal modified the request, but did not
affect the initial no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 13, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: October 15, 1998, as
supplemented February 3, and February 12, 1999.
Brief description of amendment: The amendment authorizes a revision
to the TMI-1 updated final safety analysis report (UFSAR) for use of
revised atmospheric dispersion factors (X/Q) (obtained by utilizing
recent meteorological data) in determining Chapter 14 postulated
accident analysis radiological dose consequences at Technical
Specification Section 5.1.1 defined exclusion area boundary (EAB) and
low population zone (LPZ).
Date of issuance: April 15, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 210.
Facility Operating License No. DPR-50. Amendment authorizes changes
to the UFSAR.
Date of initial notice in Federal Register: November 18, 1999 (63
FR 64117).
The February 3, and February 12, 1999, letters were within the
scope of the original application and did not change the staff's no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 15, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: December 28, 1998, as
supplemented March 1 and 29, 1999.
Brief description of amendment: The amendment revises Technical
Specification (TS) 2.2.1, ``Limiting Safety System Settings-Reactor
Trip Setpoints,'' to reflect revised loss of normal feedwater flow
analyses.
Date of issuance: April 8, 1999.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 232.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6701). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, Attn: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: January 18, 1999.
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.6.1.2, ``Containment Systems--Containment
Leakage,'' and also revises the related TS bases and Final Safety
Analysis Report sections. The revisions relate to changes in the
secondary containment bypass leakage.
Date of issuance: April 14, 1999.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 234.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6703). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated April 14, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, Attn: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: February 10, 1999.
Brief description of amendment: The amendment incorporates
alternative inspection requirements into Technical Specification
Surveillance Requirement 3/4.4.10, ``Structural Integrity,'' for the
reactor coolant pump flywheel.
Date of issuance: April 16, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment No.: 169.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 10, 1999 (64 FR
11964).
[[Page 24209]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 16, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, Attn: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: February 5, 1999, as
supplemented March 1, 1999.
Brief description of amendments: The amendments revise certain
requirements for repair of defective steam generator tubs specified in
Technical Specification 4.12, ``Steam Generator Tube Surveillance,''
based on the latest revision to a previously approved methodology.
Date of issuance: April 15, 1999.
Effective date: April 15, 1999, with full implementation within 30
days.
Amendment Nos.: 144 Unit 1--135 Unit 2.
Facility Operating License Nos. DPR-42 and DPR-60. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 10, 1999 (64 FR
11964). The March 1, 1999, supplement provided corrected Technical
Specification pages. This information was within the scope of the
original Federal Register notice and did not change the staff's initial
proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 15, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No.1, Washington County, Nebraska
Date of amendment request: March 18, 1998.
Brief description of amendment: The amendment revises Technical
Specification (TS) 5.2.f and TS 5.11.2 to change the title of ``Shift
Supervisor'' to ``Shift Manager.''
Date of issuance: April 15, 1999.
Effective date: April 15, 1999.
Amendment No.: 190.
Facility Operating License No. DPR-40. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 8, 1998 (63 FR
17227).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 15, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: May 16, 1996.
Brief description of amendment: The amendment revises requirements
for Plant Operating Review Committee review of fire protection program
and procedure changes.
Date of issuance: April 12, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 252.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34895).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 12, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of application for amendment: December 16, 1998, as
supplemented March 22, 1999.
Brief description of amendment: This amendment revised Technical
Specification (TS) Surveillance Requirements 4.8.1.1.2 and 4.8.1.1.3,
Table 4.8.1.1.2-1, and the associated Bases. These changes removed the
emergency diesel generator accelerated testing and special reporting
requirements from the TSs in accordance with the guidance provided in
Generic Letter 94-01.
Date of issuance: April 14, 1999.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 119.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 13, 1999 (64 FR
2251).
The supplemental letters provided clarifying information that did
not change the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 14, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of application for amendment: June 12, 1998, as supplemented
July 23, 1998 and September 8, 1998.
Brief description of amendment: The amendment revises Technical
Specification (TS) Limiting Condition for Operation Sections 3.7.1.1,
3.7.1.2, and 3.7.1.3. Specifically, the changes revise the Ultimate
Heat Sink limits for river water temperature, in order to increase
operational flexibility. In addition, the Station Service Water System
(SSWS) and Safety Auxiliaries Cooling System (SACS) TS Action
Statements have been revised to provide additional restrictions on
continued plant operation. These revisions provide more explicit TS
direction for plant operation under limiting SSWS/SACS configurations.
Date of issuance: April 19, 1999.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 120.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 1, 1999 (63 FR
35995) The July 23, 1998, and September 8, 1998, supplements provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination or expand the scope of
the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 19, 1999.
No significant hazards consideration comments received: No
[[Page 24210]]
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
South Carolina Electric & Gas Company, South Carolina Public Service
Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station, Unit
No. 1, Fairfield County, South Carolina
Date of application for amendment: September 18, 1998, as
supplemented by letter dated February 5, 1999.
Brief description of amendment: The amendment revises Virgil C.
Summer Nuclear Station Technical Specifications to permit use of the
BEACON system. BEACON is a core power distribution monitoring and
support system based on a three-dimensional nodal code.
Date of issuance: April 9, 1999.
Effective date: April 9, 1999.
Amendment No.: 142.
Facility Operating License No. NPF-12: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64121).
The February 5, 1999, submittal contained clarifying information
only, and did not change the initial no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated April 9, 1999.
No significant hazards consideration comments received: No
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San
Diego County, California
Date of application for amendments: January 24, 1997.
Brief description of amendments: The amendments revised
Surveillance Requirement (SR) 3.8.1.9 to Technical Specification 3.8.1,
``AC Sources--Operating,'' to more accurately reflect test conditions
and plant design requirements.
Date of issuance: April 9, 1999.
Effective date: April 9, 1999, to be implemented within 30 days
from the date of issuance.
Amendment Nos.: Unit 2-151; Unit 3-143.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 11, 1998 (63
FR 6997).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 9, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: January 20, 1999.
Brief description of amendments: The amendments revise the
descriptive details of Technical Specification 4.7.1.2.1.a, regarding
performance testing of the Auxiliary Feedwater (AFW) pumps, to more
closely adhere to NUREG-1431, ``Improved Standard Technical
Specifications for Westinghouse Plants.'' This involves relocating the
surveillance-required numerical values for the AFW pump performance
test discharge pressure and flow rate to the South Texas Project
Updated Final Safety Analysis Report.
Date of issuance: April 16, 1999.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment Nos.: Unit 1--Amendment No. 105; Unit 2--Amendment No.
92.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 24, 1999 (64
FR 9201).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 16, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: January 26, 1999.
Brief description of amendments: The amendments revise part of the
inservice inspection requirements for the reactor coolant pump flywheel
from an in-place ultrasonic volumetric examination of the areas of
higher stress concentration at the bore and keyway at approximately 3-
year intervals and a surface examination of all exposed surfaces and
complete ultrasonic volumetric examination at approximately 10-year
intervals to ultrasonic examination over the volume from the inner bore
of the flywheel to the circle of one-half the outer radius once every
10 years.
Date of issuance: April 16, 1999.
Effective date: April 16, 1999, to be implemented within 30 days of
issuance.
Amendment Nos.: Unit 1--Amendment No. 106; Unit 2--Amendment No.
93.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 10, 1999 (64 FR
11968).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 16, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: September 30, 1998.
Brief description of amendments: Revises Units 1 and 2 Technical
Specification (TS) Section 3/4.4.5, ``Steam Generator'' Surveillance
Requirements. The future installation of the new Delta 94 steam
generators at the South Texas Project, Units 1 and 2 necessitates
changes to the steam generator tube sample selection and inspection
requirements; inservice inspection frequencies; acceptance criteria;
and inspection reporting requirements.
Date of issuance: April 19, 1999.
Effective date: April 19, 1999, to be implemented following the
replacement of Unit 1 Model E steam generators with Model delta94 steam
generators and prior to Unit 1 operation with the delta94 steam
generators installed.
Amendment Nos.: Unit 1--Amendment No. 107; Unit 2--Amendment No.
94.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 4, 1998 (63 FR
59595).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 19, 1999.
No significant hazards consideration comments received: No.
[[Page 24211]]
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 6, 1997, as supplemented by
letters dated September 4 and 18, 1997, December 9, 1997, and February
4, 1999.
Brief description of amendments: The amendments revise Technical
Specification (TS) Table 2.2-1 and TS 3/4.2.5 to allow the reactor
coolant system total flow rate to be determined using cold leg elbow
tap differential pressure measurements.
Date of issuance: April 19, 1999.
Effective date: As of its date of issuance to be implemented within
7 days of issuance.
Amendment Nos.: Unit 1--Amendment No. 108; Unit 2--Amendment No.
95.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 14, 1997 (62 FR
43556).
The September 4 and 18, 1997, December 9, 1997, and February 4,
1999, letters provided clarifying information that did not change the
original application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 19, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,
Unit 1, Hamilton County, Tennessee
Date of application for amendments: August 27, 1998, supplemented
by letter dated March 19, 1999 (TS 98-04).
Brief description of amendments: The amendments change the
Technical Specifications (TS) for Sequoyah Nuclear Plant, Unit 2
reactor by adding a sentence at the end of TS Section 5.3 authorizing
installation of a limited number of lead test assemblies containing
downblended uranium in accordance with Topical Report BAW-2328.
Date of issuance: April 12, 1999.
Effective date: April 12, 1999.
Amendment Nos.: 234.
Facility Operating License No. DPR-79: The amendment revises the
TS.
Date of initial notice in Federal Register: March 10, 1999 (64 FR
11969). The supplemental letter of March 19, 1999 did not change the
initial proposed no significant hazards condition determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 12, 1999.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: June 29, 1998, as supplemented
by letter dated February 19, 1999.
Brief description of amendment: The amendment revised Technical
Specification (TS) 3.7.1.7 operability requirements to require four
atmospheric steam dump (ASD) lines to be operable. Other changes were
made to TS 3.7.1.7 to address action statements and surveillance
requirements for the four ASD lines.
Date of issuance: April 20, 1999.
Effective date: April 20, 1999, to be implemented within 30 days
from the date of issuance.
Amendment No.: 131.
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48271).
The February 19, 1999, supplemental letter provided additional
clarifying information, did not expand the scope of the application as
originally noticed, and did not change the staff's original proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 20, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Elmer Ellis Library,
University of Missouri, Columbia, Missouri 65201.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: May 28, 1998, as supplemented
December 11, 1998.
Brief description of amendments: These amendments revise Technical
Specifications (TS) to provide a specific numerical setting for reactor
trip, reactor coolant pump trip, and auxiliary feedwater initiation on
a loss of power to the 4 kilovolt (kV) buses. Changes to the bases for
the affected TS sections are also being made.
Date of issuance: April 23, 1999.
Effective date: April 23, 1999.
Amendment Nos.: Unit 1-189; Unit 2-194.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 15, 1998 (63 FR
38208).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 23, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Dated at Rockville, Maryland, this 28th day of April 1999.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 99-11119 Filed 5-4-99; 8:45 am]
BILLING CODE 7590-01-P