[Federal Register Volume 63, Number 231 (Wednesday, December 2, 1998)]
[Notices]
[Pages 66590-66609]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-31931]
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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 November 6, 1998, through November 19, 1998.
The last biweekly notice was published on November 18, 1998 (63 FR
64106).
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.
[[Page 66591]]
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 January 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 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.
Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414,
Catawba Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: November 11, 1998.
[[Page 66592]]
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) to correct Surveillance
Requirements (SRs) 3.6.11.6 and 3.6.11.7 and the associated Bases.
These SRs currently are incorrect and do not reflect the Containment
Pressure Control System (CPCS) as designed. Therefore, the proposed
amendments would only revise the SRs; no change to the CPCS design is
involved.
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:
First Standard
Implementation of this amendment would not involve a significant
increase in the probability or consequences of an accident
previously evaluated. Approval of this amendment will have no
significant effect on accident probabilities or consequences.
The CPCS is not an accident initiating system; therefore, there
will be no impact on any accident probabilities by the approval of
this amendment. The design of the CPCS is not being modified by this
proposed amendment. The amendment merely aligns [TS] surveillance
requirements with the existing design and function of the system.
Therefore, there will be no impact on any accident consequences.
Second Standard
Implementation of this amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated. No new accident causal mechanisms are created
as a result of NRC approval of this amendment request. No changes
are being made to the plant which will introduce any new accident
causal mechanisms. This amendment request does not impact any plant
systems that are accident initiators, since the CPCS is an accident
mitigating system.
Third Standard
Implementation of this amendment would not involve a significant
reduction in a margin of safety. Margin of safety is related to the
confidence in the ability of the fission product barriers to perform
their design functions during and following an accident situation.
These barriers include the fuel cladding, the reactor coolant
system, and the containment system. The performance of these fission
product barriers will not be impacted by implementation of this
proposed amendment. The CPCS is already capable of performing as
designed. No safety margins will be 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: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Attorney for licensee: Mr. Paul R. Newton, Legal Department
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,
North Carolina.
NRC Project Director: Herbert N. Berkow.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of amendment request: October 15, 1998.
Description of amendment request: The proposed amendments would
revise the pressure-temperature limits in the Technical Specifications
for Units 1, 2, and 3. The proposed amendments would revise the heatup,
cooldown, and inservice test limitations for the reactor coolant system
of each unit to a maximum of 26 effective full-power years. The
proposed amendments would also revise the Technical Specification for
low temperature overpressure protection to reflect the revised
pressure-temperature limits of the reactor vessels.
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. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
NO.
Each accident analysis addressed in the Oconee UFSAR [Updated
Final Safety Analysis Report] has been examined with respect to the
changes to the Reactor Pressure Vessel (RPV) pressure-temperature
limit curves and related Low Temperature Overpressure settings. The
probability of any design basis accident (DBA) is not affected by
this change, nor are the consequences of a DBA affected by this
change. The revised pressure-temperature limits, which were
developed based on NRC approved methodology or ASME Code [American
Society of Mechanical Engineers Boiler and Pressure Vessel Code]
Case N-514 as described in the Technical Justification, are not
considered to be an initiator or contributor to any accident
analysis addressed in the Oconee UFSAR. The added requirement to
deactivate one pressurizer heater bank during low temperature
operation does not significantly change the probability or
consequence of any accident previously analyzed. No existing
Technical Specification requirements are being deleted with this
revision.
B. Create the possibility of a new or different kind of accident
from the accident previously evaluated?
NO.
This license amendment revises Oconee RPV pressure-temperature
limits. The revised pressure-temperature limits were developed based
on NRC approved methodology or ASME Code Case N-514 as described in
the Technical Justification. Operation of Oconee in accordance with
these proposed new Technial Specifications will not create any
failure modes not bounded by previously evaluated accidents.
Consequently, this change will not create the possibility of a new
or different accident from any accident previously evaluated.
C. Involve a significant reduction in a margin of safety?
NO.
This license amendment revises Oconee RPV pressure-temperature
limits. The revised pressure-temperature limits were developed based
on NRC approved methodology or ASME Code Case N-514 as described in
the Technical Justification. The purpose of this license amendment
is to assure that sufficient operating margin to safety is
maintained in the operation of the Oconee reactor pressure vessels
by establishing new, more limiting pressure-temperature limit curves
and adding the requirement to deactivate one pressurizer heater
bank. No plant safety limits, set points, or design parameters are
adversely affected. The fuel, fuel cladding, and Reactor Coolant
System are not impacted. Therefore, there will be no significant
reduction in any margin of safety.
Duke [Duke Energy Corporation] has concluded based on this
information that there are no significant hazards considerations
involved in this amendment request.
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: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC.
NRC Project Director: Herbert N. Berkow. Duquesne Light Company, et
al., Docket No. 50-334, Beaver Valley Power Station, Unit No. 1,
Shippingport, Pennsylvania
Date of amendment request: November 11, 1998.
Description of amendment request: The proposed amendment would
modify License Condition 2.C(9) to allow, on a one time only basis, an
extension to the steam generator inspection interval of technical
specification surveillance 4.4.5.3.b. This
[[Page 66593]]
would allow the steam generator inspection interval to coincide with
the 13th refueling outage or the end of 500 effective full power days,
whichever occurs sooner.
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?
The proposed change is temporary and allows a one time extension
of specific surveillance requirements for Cycle 13 to allow
surveillance testing to coincide with the 13th (1R13) refueling
outage. The proposed surveillance interval extension will not cause
a significant reduction in system reliability nor affect the ability
of a system to perform its design function. Current monitoring of
plant conditions and the surveillance monitoring required during
normal plant operation will be performed as usual to assure
conformance with technical specification operability requirements.
The technical specification steam generator tube inspection is
intended to prevent the Steam Generator Tube Rupture analyzed in
[Updated Final Safety Analysis Report] UFSAR Section 14.2.4 by
maintenance of the integrity of the primary to secondary coolant
boundary represented by steam generator tubes. The process by which
this integrity is maintained is inspection of steam generator tubes
at prescribed intervals, and the removal of defective tubes from
service. Inspection intervals are based on preventing corrosion
growth from exceeding tube structural limits, thereby preventing
tube failure. The 1997 steam generator inspection characterized
existing steam generator tube degradation, and degraded tubes were
removed from service at that time. Degradation growth rates were
evaluated for the next operating interval and it was determined that
the steam generator tube structural integrity is maintained.
Degradation of steam generator tubes was prevented during the
extended outage by a carefully controlled, corrosion prevention
program.
The proposed change does not affect the UFSAR and is consistent
with changes granted for other plants. The surveillance extension
does not involve a change to plant equipment and does not affect the
performance of plant equipment used to mitigate an accident. This
change, therefore, does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Extending the surveillance interval for the performance of
specific inspections will not create the possibility of any new or
different kind of accidents. No change is required to any system
configurations, plant equipment or analyses.
Steam generator tube inspections determine tube integrity and
provide reasonable assurance that a tube rupture or primary to
secondary leak will not occur. Accidents involving steam generator
tube rupture are analyzed in UFSAR Section 14.2.4, ``Steam Generator
Tube Rupture.'' The only type of accident that can be postulated
from extending the steam generator inspection interval would be a
tube leak or rupture which are analyzed in the UFSAR. No new failure
modes are created by the surveillance extension. Therefore, this
change will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
Surveillance interval extensions will not impact any plant
safety analyses since the assumptions used will remain unchanged.
The safety limits assumed in the accident analyses and the design
function of the equipment required to mitigate the consequences of
any postulated accidents will not be changed since only the
surveillance interval is being extended. Based on engineering
judgement, extending the surveillance interval for the performance
of these specific inspections does not involve a significant
reduction in the margin of safety derived from the required
surveillances.
The margin of safety depends upon maintenance of specific
operating parameters within design limits. In the case of steam
generators, that margin is maintained through assurance of tube
integrity as the primary to secondary boundary. Assurance of tube
integrity is provided through periodic in-service inspection of
tubes and removal of defective tubes from service. Additional margin
is provided through protection from possible consequences of steam
generator tube failure by mitigation systems. Radiation monitors
provide a detection capability of primary to secondary leakage to
enable a prompt response. Maintenance of the steam generator water
chemistry in accordance with [Electric Power Research Institute]
EPRI guidelines provides additional margin of safety. Therefore, the
plant will be maintained within the analyzed limits and the proposed
extension will not significantly reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Attorney for Licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Robert A. Capra Entergy Operations, Inc.,
Docket No. 50-368, Arkansas Nuclear One, Unit No. 2, Pope County,
Arkansas
Date of amendment request: June 30, 1998.
Description of amendment request: The proposed change modifies the
Engineered Safety Features Actuation System (ESFAS) portion of the
Arkansas Nuclear One, Unit-2 (ANO-2) Plant Protection System (PPS).
This modification is designed to defeat the backup power supply for the
auctioneered power sources for channel A and D Reactor Protective
System (RPS) and ESFAS bistables, and to provide selective logic for
Emergency Feedwater Actuation Signals and Main Steam Isolation Signals.
This will ensure that ESFAS will have the redundancy and independence
sufficient to assure that (1) no single failure results in loss of the
protection function with a channel in indefinite bypass, and (2)
removal from service of any component or channel does not result in
loss of the required minimum redundancy required by the ANO-2 Technical
Specifications (TSs). The proposed modification to the ANO-2 PPS has
been determined to involve an Unreviewed Safety Question in accordance
with 10 CFR 50.59(a)(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:
An evaluation of the proposed change has been performed in
accordance with 10 CFR 50.91(a)(1) regarding no significant hazards
considerations using the standards in 10 CFR 50.92(c). A discussion
of these standards as they relate to this amendment request follows:
Criterion 1--Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The ANO-2 Plant Protection System (PPS) includes the electrical
and mechanical devices and circuitry (from sensors to actuation
device input terminals) involved in generating signals associated
with the two protective functions, Engineered Safety Feature
Actuation System (ESFAS) and Reactor Protective System (RPS). The
RPS is that portion of the PPS which generates signals that actuate
a reactor trip. The ESFAS is that portion of the PPS which generates
signals that actuate Engineered Safety Features (ESF) to mitigate
the consequences of an accident.
The ANO-2 Safety Analysis Report (SAR) section 15.1.31 ``Loss Of
One DC System'' analyzes failure of a DC bus (FODCB) as initiator
and its causes. The causes for the FODCB are DC leg to leg fault in
the bus or in the power distribution circuit from the battery. Since
the proposed change has no impact on the accident initiator, the
frequency of occurrence is not changed. In order for the FODCB as a
single failure with an accident to de-energize two [Vital Instrument
Buses (]VIBs[)], the FODCB would have to occur prior to the safety
bus
[[Page 66594]]
energization by offsite bus fast transfer or prior to safety bus
energization by the emergency diesel generator (EDG). The potential
for de-energization of one pair of VIBs is, therefore, limited to
time from initiation of the accident to time for safety bus response
to the secondary plant and Reactor Protective System trips.
The effects of the FODCB are being revised to assume a secondary
plant trip that results in de-energization of one power division.
The existing analysis conclusions remain unchanged. The accident
analysis is being revised to include de-energization of a pair of
vital AC instrument channels. De-energization of two vital AC
sources has not been previously documented as a design bases event.
Auctioneered bistable power supplies for Plant Protection System
(PPS) channels A and D are being modified to a single power source
for each of these two channels. Single channel trips will result for
all PPS functions in channels A or D for loss of its single channel
bistable power source. The PPS channels B and C auctioneered power
supplies remain unchanged to maintain Recirculation Actuation Signal
(RAS) response to a FODCB.
Regarding PPS measurement channels with increasing signal
setpoints, de-energization of a single power supply either results
in failure of a measurement channel (B or C) to a non-tripped state
or in failure of a measurement channel (A or D) to a tripped state.
Neither single channel failure scenario impacts accident initiation
or mitigation. For PPS measurement channels with decreasing signal
setpoints the single channel de-energization events result in
failure of a single affected measurement channel to a tripped state.
The PPS two out of three logic design with a channel bypassed
ensures operability with a single channel failure. Neither condition
impacts accident frequency or consequences.
With the exception of Recirculation Actuation Signal (RAS) and
Emergency Feedwater Actuation Signal (EFAS), a FODCB results in an
automatic ESFAS initiation for those functions with decreasing
signal setpoints. For other ESFAS functions with a decreasing
signal, channels A and C or channels B and D fail to the tripped
state. For those functions with an increasing signal setpoint
(including EFAS), a FODCB results in a single channel failing not
tripped, one channel tripping, and two channels remaining
functional. System level functions remain operable with either a one
out of two logic (no channels bypassed) or a one out of one logic
(with a channel bypassed).
Interposing relay actuation logic has changed from single trip
path to selective trip path logic. This change insures emergency
feedwater (EFW) discharge valves will receive an automatic open or
close demand based on steam generator level and pressure demands.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
Criterion 2--Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
In response to de-energization of a pair of Vital Instrument
Buses (VIBs), those ESFAS functions with increasing signal
setpoints, as a minimum, remain functional with one out of one
logic. One channel trips, one channel does not trip, and two
channels remain functional. One of the functional channels may be
bypassed without impact on operability. The trip response of those
ESFAS functions with decreasing signal to trip setpoints remains
unchanged.
EFAS coincidence logic to close the EFW discharge valves
requires three out of four channels to be in a non-tripped state.
With a FODCB one channel is tripped, one channel is not tripped, and
two channels are functional. The close logic becomes two out of two
with a FODCB.
By defeating the auctioneered bistable power sources for PPS
channel A and D bistables, PPS measurement channel A or D will fail
to its tripped state. This change ensures no more than one channel
(B or C) fails to a non-tripped state for the FODCB.
With selective logic EFAS pump discharge valves will receive
control signals to initiate emergency feedwater and to terminate
emergency feedwater flow by open and close demands generated
independent of the 120 Volt channel pair de-energization.
The existing ANO-2 Failure Modes and Effects Analysis does not
document failure of a pair of vital instrument AC channels. Neither
the 120 Volts AC nor the 125 Volt DC system single failure analysis
assumes failure of two channels of 120 Volts AC. Even though the
failure of either pair of VIBs caused by a FODCB is not a result of
the proposed change, the SAR change will address the potential for
de-energization of a pair of instrument buses. The ANO-2 SAR will be
updated to reflect the documentation and modification of the PPS
design to ensure safe plant response.
Even though the plant response to FODCB is being modified, the
proposed ANO-2 PPS design resolution does not create the possibility
of a new or different kind of accident from any previously evaluated
in the SAR. The PPS will have the redundancy and independence
sufficient to assure that (1) no single failure results in loss of
the protection function, and (2) removal from service of any
component or channel does not result in loss of the required minimum
redundancy required by the TS. PPS will also meet the single failure
criterion of IEEE 279-1971 to the extent that any single failure
within the system does not prevent proper protective action at the
system level and no single failure will defeat more than one of the
four protective channels associated with any one trip function.
Criterion 3--Does Not Involve a Significant Reduction in the
Margin of Safety.
Technical Specification Bases 3/4.3.1 & 3/4.3.2 assure
sufficient PPS redundancy is maintained to permit a channel to be
bypassed. Under the current design, a FODCB will result in reduction
of margin by decreasing the number of functional channels to less
than two. However, with the proposed modification removal from
service of any component or channel for indefinite bypass will not
result in loss of the minimum redundancy required by the TS. This
activity will restore the margin by ensuring ESFAS required
functions remain capable of automatic actuation with a FODCB.
Therefore, this change does not involve a significant reduction
in the margin of safety.
Based upon the reasoning presented above and the previous
discussion of the amendment request, Entergy Operations has
determined that even though the proposed PPS design description
results in an accident or malfunction of a different type, the
requested change does not involve a significant hazards
consideration.
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: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502.
NRC Project Director: John N. Hannon.
Florida Power and Light Company, et al., Docket No. 50-335, St. Lucie
Plant, Unit No. 1, St. Lucie County, Florida
Date of amendment request: October 29, 1998.
Description of amendment request: The proposed amendment would
revise the terminology used in the St. Lucie Plant Technical
Specifications (TS) relative to the implementation and automatic
removal of certain reactor protection system trip bypasses to ensure
that the meaning of explicit terms used in the TS are consistent with
the intent of the stated requirements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments are administrative in nature, and do not
change the function or the setpoints of the RPS trip bypass
features. The revisions simply make corrections to the Notation of
TS Tables 2.2-1 and 3.3-1 to ensure that the meaning of explicit
terms used in the Notes is consistent with the intent of the stated
requirements based on the St. Lucie plant design. The proposed
technical specification changes do not involve accident initiators,
do not change the configuration or method of operation of any plant
equipment that is used to mitigate
[[Page 66595]]
the consequences of an accident, and do not alter any conditions
assumed in the plant accident analyses. Therefore, operation of
either facility in accordance with its proposed amendment would not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendments are administrative in nature and will
not change the physical plant or the modes of plant operation
defined in the facility operating licenses. The changes do not
involve the addition or modification of equipment nor do they alter
the design or operation of plant systems. Therefore, operation of
either facility in accordance with its proposed amendment would not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The proposed amendments are administrative in nature and do not
change the function or the setpoints of the RPS trip bypass
features. The revisions simply make corrections to the Notation of
TS Tables 2.2-1 and 3.3-1 to ensure that the meaning of explicit
terms used in the Notes is consistent with the intent of the stated
requirements based on the St. Lucie plant design. The proposed
changes do not alter the basis for any technical specification that
is related to the establishment of, or the maintenance of, a nuclear
safety margin. Therefore, operation of either facility in accordance
with its proposed amendment would 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. This
notice is intended to replace an exigent notice of consideration of
issuance of amendment for St. Lucie Unit 1, previously published as
exigent TS amendments for both St. Lucie Units 1 and 2 in the Federal
Register (63 FR 59809). The amendment request for St. Lucie Unit 2 will
continue to be considered as an exigent amendment as noticed in the
Federal Register (63 FR 59809).
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Frederick J. Hebdon.
GPU Nuclear, Inc, et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of amendment request: November 10, 1998.
Description of amendment request: The proposed Technical
Specification (TS) change would remove the restriction on the sale or
lease of property within the exclusion area and replace the restriction
with a requirement to retain complete authority to determine and
maintain sufficient control of all activities including the authority
to exclude or remove personnel and property within the minimum
exclusion distance. A TS Bases page for the proposed change is
included. Also included are clarifications and administrative changes
which (1) clarify TS definition 1.38 to become ``Site Boundry'' from
the current term ``Exclusion Area'' to be consistent with 10 CFR
20.1003 definition for Site Boundry and the 10 CFR 100.3 definition of
Exclusion Area, (2) convert the one occurrence of the use of TS
definition from Exclusion Area to Site Boundry in TS 6.8.4(a)(9), and
(3) revise and update the Table of Contents for Section I Definitions.`
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would operation of the facility in accordance with the
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
The proposed change is administrative in nature and does not
affect the purpose, function, performance, operability or testing of
and does not make any physical or procedural changes to plant
systems, structures or components. Also, all existing technical
specification limiting conditions for operation and surveillance
requirements are retained.
[Technical Specification Change Request] TSCR 264 does not
change the size or location of the exclusion area. Since the
exclusion area size and location are not being changed and no
physical or procedural changes are being made to the plant,
radiological consequences in the exclusion area are not affected by
this TSCR.
This change addresses the existing technical specification
restriction on the sale or lease of property within the ``exclusion
area'' by ensuring that the licensee will retain at all times the
complete authority to determine and maintain sufficient control of
all activities through ownership, easement, contract and/or other
legal instruments on property within the minimum exclusion distance
including the authority to exclude or remove personnel and property
within the minimum exclusion distance.
Therefore, since no physical or procedural changes are being
made to existing plant systems, structures or components and since
the proposed change requires the licensee to retain complete
authority and sufficient control of all activities in the exclusion
area, operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Would operation of the facility in accordance with the
proposed change create the possibility of a new or different kind of
accident from any accident previously evaluated?
The p[ro]posed change is administrative in nature and does not
affect the purpose, function, performance, operability or testing of
and does not make any physical or procedural changes to plant
systems, structures or components. Also, all existing technical
specification limiting conditions for operation and surveillance
requirements are retained.
This change addresses the existing technical specification
restriction on the sale or lease of property within the ``exclusion
area'' by ensuring that the licensee will retain at all times the
complete authority to determine and maintain sufficient control of
all activities through ownership, easement, contract and/or other
legal instruments on property within the minimum exclusion distance
including the authority to exclude or remove personnel and property
within the minimum exclusion distance.
Therefore, since no physical or procedural changes are being
made to existing plant systems, structures or components and since
the proposed change requires the licensee to retain complete
authority and sufficient control of all activities in the exclusion
area, operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. Would operation of the facility in accordance with the
proposed change involve a significant reduction in a margin of
safety?
The p[ro]posed change is administrative in nature and does not
affect the purpose, function, performance, operability or testing of
and does not make any physical or procedural changes to plant
systems, structures or components. Also, all existing technical
specification limiting conditions for operation and surveillance
requirements are retained.
This change addresses the existing technical specification
restriction on the sale or lease of property within the ``exclusion
area'' by ensuring that the licensee will retain at all times the
complete authority to determine and maintain sufficient control of
all activities through ownership, easement, contract and/or other
legal instruments on property within the minimum exclusion distance
including the authority to exclude or remove personnel and property
within the minimum exclusion distance.
Therefore, since no physical or procedural changes are being
made to existing plant
[[Page 66596]]
systems, structures or components and since the proposed change
requires the licensee to retain complete authority and sufficient
control of all activities in the exclusion area, operation of the
facility in accordance with the proposed amendment will not involve
a significant reduction in margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cecil O. Thomas.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point
Nuclear Station, Unit 2 (NMP2), Oswego County, New York
Date of amendment request: October 16, 1998.
Description of amendment request: The proposed amendment would make
the following revisions to Technical Specifications (TSs) 3/4.7.1.1:
(1) Ensure that four service water (SW) pumps are operating with the
divisional cross connect valves open during Operational Condition 1, 2
and 3 (current TS requires two SW pumps associated with one loop to be
operating); (2) Increase the number of division 1 and 2 heaters
required to be operable from 7 per division per intake to 14 per
division per intake; (3) The actions necessary for having less than the
required equipment is being revised to reflect the new limits for SW
equipment; and (4) SW supply header discharge water temperature is
being increased from 81 to 82 deg.F. TS 3.7.1.2, Table 3.3.9-1, and
Table 4.3.9.1-1 are revised to add ``when handling irradiated fuel in
the secondary containment'' to the applicability section. Table 3.3.9-1
is being revised to decrease the temperature at which the Intake
Deicing Heaters are required to be in service from 39 to 38 degrees F.
TS 3.7.1.2 proposed change is to specify that the necessary portions of
the SW system needed to support equipment required to be operable shall
be operable; the Action Section proposed revision reflects this change.
TS 4.7.1.2.1 surveillance requirement proposed change is to increase
the flow rate of SW pumps from 6500 GPM to 9000 GPM and to change the
SW pumps pressure from 80 psi discharge pressure to 70 psi differential
pressure; TS 4.7.1.2.2 is being revised to decrease the intake tunnel
water temperature from 39 to 38 degrees F. The surveillance for the
Intake Deicing Heaters is being changed to reflect the increase in the
number of heaters required. The title of ``Plant Service Water System''
is being changed to ``Service Water System.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The SW System is a once-through system which supplies water from
Lake Ontario to various essential and non-essential components, as
required, during normal plant operation and shutdown conditions. The
System is designed with suitable redundancy to provide a reliable
source of cooling water for the removal of heat from essential plant
components, including the RHR [residual heat removal] heat
exchangers, the EDGs [emergency diesel generators], and room coolers
for ECCS [emergency core cooling system] equipment, which are
required for safe reactor shutdown following a LOCA.
LCO 3.7.1.1 and LCO 3.7.1.2 each currently requires two
independent SW System loops to be operable, with one of the loops in
operation. The current LCOs do not provide adequate guidance
regarding the minimum number of operating pumps. NMPC [Niagara
Mohawk Power Corporation] proposes to revise LCO 3.7.1.1 and its
associated Actions and SRs to provide assurance that four SW pumps
are operable and are operating within acceptable system parameters,
with the divisional cross-connect valves open, during Operational
Conditions 1, 2, and 3 to meet the limiting LOCA analysis
assumptions.
TS Section 3/4.7.1 currently specifies a maximum SW supply
header discharge water temperature of 81 degrees F and a limiting
temperature for Intake Deicing Heater ystem operability (intake
water) temperature of 39 degrees F. In addition, TS Table 3.3.9-1,
Action 144, requires the Intake Deicing Heater System heaters to be
placed in service when the Lake Ontario water temperature reaches 39
degrees F. NMPC proposes to revise Action 144 of TS Table 3.3.9-1
and TS LCO 3.7.1.1, including its associated Actions and SRs
[surveillance requirements], to increase the supply header discharge
water temperature to its analytical limit of 82 degrees F and reduce
the limiting temperature for the Intake Deicing Heater System Action
and operability requirements to 38 degrees F.
Appropriate changes to LCO 3.7.1.2 and its associated Actions
and SRs are also proposed in order to assure consistency with the SW
System analyses assumptions during shutdown conditions. The current
LCO Actions do not account for the varying flows and heat loads that
may be required for various plant shutdown conditions. The revision
to the Applicability for LCO 3.7.1.2 and TS Tables 3.3.9-1 and
4.3.9.1-1 will assure that the SW System is operable during periods
when irradiated fuel is being handled in the secondary containment
and essential loads cooled by the SW System are required to be
operable (e.g., EDG). A footnote has been added to define
Operational Condition * and is consistent with similar footnotes in
the TSs. The proposed changes will assure that the necessary ortions
of the SW System and the necessary Divisions of the Intake Deicing
Heater System heaters are operable that are supporting equipment
required to be operable.
It is further proposed to change the system title identified in
the Index and in TS Section 3/4.7.1, including the LCOs and SRs,
from ``Plant Service Water System'' to ``Service Water System'' to
be consistent with the NMP2 [Nine Mile Unit 2] UFSAR [Updated Final
Safety Analysis Report].
The changes do not involve any physical alteration of the plant,
and the SW System will remain capable of providing sufficient
cooling flow for the essential cooling loads during plant operation
and also during plant shutdown. The changes will have no impact on
the design or function of the SW System and its components, thus
assuring that the characteristics and functional performance are
maintained consistent with the event precursors and the conditions
and assumptions of the current design basis accident and transient
analyses. The changes to the LCO AOTs [allowed outage times] are
either consistent with or are more conservative than the current
AOTs. Based on the above, adequate assurance is provided that the
probability of event initiation will remain as previously analyzed.
Maintaining four pumps operating within acceptable system
parameters, with the divisional cross connect valves open, during
Operational Conditions 1, 2, and 3 provides assurance that the
essential functions supported by the SW System are maintained.
Particularly, adequate SW flow assures that the primary and
secondary containments can perform their intended functions of
limiting the release of radioactive materials to the environment
following a LOCA. The small (1 degree F) change in the SW supply
header discharge water (UHS) temperature and Intake Deicing Heater
System actuation temperature maintain the current design basis for
the UHS and SW Systems such that there will be no impact on the LOCA
analyses assumptions or conclusions. The proposed changes to the SW
System TSs do not adversely affect the capability of plant systems,
structures, and components to respond to any accident in Operational
Conditions 4, 5, and *. As a result, there will be no degradation of
the primary or secondary containment or any other fission product
barriers which could increase the radiological consequences of an
accident. In addition, other essential accident mitigation equipment
supported by the SW System will not be adversely impacted. It is,
therefore,
[[Page 66597]]
concluded that operation of NMP2, in accordance with the proposed
amendment, will not involve a significant increase in the
probability or 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 changes do not result in any hardware changes or physical
alteration of the plant which could introduce new equipment failure
modes, and there will be no impact on the design or function of the
SW System or its components. The primary and secondary containment
post-LOCA responses remain within previously assessed limits of
temperature and pressure. Furthermore, adequate cooling flow is
assured during plant operation and also during shutdown conditions
such that essential systems and components remain within their
applicable design limits. It is, therefore, concluded that no
requirements are eliminated or new requirements imposed which could
affect equipment or plant operation such that new credible accidents
are introduced. Accordingly, operation of NMP2, 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 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 changes provide assurance that the SW System will remain
capable of providing sufficient cooling flow for the essential
cooling loads during plant operation and also during plant shutdown
such that essential systems and components remain within their
applicable design limits. The changes will have no impact on the
design or function of the SW System and its components, thus
assuring that the characteristics and functional performance are
maintained consistent with the conditions and assumptions of the
current design basis accident and transient analyses. Maintaining
four pumps operating within acceptable system parameters, with the
divisional cross connect valves open, during Operational Conditions
1, 2, and 3 provides assurance that post-LOCA radioactive releases
are maintained within 10 CFR 100 limits. The small (1 degree F)
change in the SW supply header discharge water (UHS) temperature and
the limiting temperature for the Intake Deicing Heater System Action
and operability requirements maintains the current design basis for
the UHS and SW Systems such that there will be no impact on the LOCA
analyses assumptions or conclusions.
These changes will not result in a reduction in margin to the
System analytical limits. Furthermore, maintaining the intake bar
surface temperature at least 1 degree F above freezing provides an
adequate margin to prevent the adherence of ice, and provides
assurance that sufficient flow area is always heated such that the
SW System will remain capable of providing adequate cooling flow in
the event of a LOCA. Similarly, maintaining the required SW System
flow and temperature during Operational Conditions 4, 5, and * will
assure that the associated equipment is operable such that
radioactive releases are maintained within 10 CFR 100 limits. It is,
therefore, concluded that the changes do not eliminate any
requirements, impose any new requirements, or alter any physical
parameters which significantly reduce the margin to an acceptance
limit or adversely affect the margins associated with the fission
product barriers as established by the design basis accident and
transient analyses. Accordingly, 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 Project Director: S. Singh Bajwa.
Northeast Nuclear Energy Company (NNECO) et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of amendment request: September 28, 1998.
Description of amendment request: The proposed amendment would
change Technical Specifications 3.3.2.1, ``Instrumentation--Engineered
Safety Features Actuation System''; 3.4.6.2, ``Reactor Coolant System--
Reactor Coolant System Leakage''; 3.4.8, ``Reactor Coolant System--
Specific Activity''; 3.6.2.1, ``Containment Systems--Depressurization
and Cooling Systems Containment Spray and Cooling Systems''; 3.6.5.1,
``Containment Systems--Secondary Containment Enclosure Building
Filtration System''; 3.7.6.1, ``Plant Systems--Control Room Emergency
Ventilation System''; and 3.9.15, ``Refueling Operations--Storage Pool
Area Ventilation System--Fuel Storage.'' Information would also be
added to the Bases of the associated Technical Specifications to
address the proposed changes.
The proposed amendment would also revise the Operating License DPR-
65 by incorporating a change to the Millstone Unit No. 2 Final Safety
Analysis Report (FSAR). The change to the FSAR is associated with the
revised main steamline break analyses, new determination of the
radiological consequences of a main steamline break, and a revised
determination of the radiological consequences of the design basis
loss-of-coolant accidents (LOCAs).
The proposed changes to the main steamline break analysis, as
described in the FSAR, are based on the revised Siemens Power
Corporation steamline break methodology. The report describing the
revised methodology was submitted by Siemens Power Corporation to the
NRC for approval in a letter dated June 30, 1998. The revised
methodology was used to perform the Millstone Unit No. 2 plant-specific
analysis for post-scram main steamline break. This plant-specific
analysis was submitted by NNECO in a letter dated August 12, 1998,
which proposed to change the list of documents in the Technical
Specifications that describe the analytical methods used to determine
the core operating limits. The proposed changes contained in this
letter assume approval of the previously submitted revised Siemens
Power Corporation steamline break methodology, and the changes to the
list of documents in the Millstone Unit No. 2 Technical Specifications
that describe the analytical methods used to determine the core
operating limits.
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:
In accordance with 10CFR50.92, NNECO has reviewed the proposed
changes and has concluded that they do not involve a significant
hazards consideration (SHC). The basis for this conclusion is that
the three criteria of 10CFR50.92(c) are not compromised. The
proposed changes do not involve an SHC because the changes would
not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Analyses Changes
The main steam line break analyses and the determinations of the
radiological consequences of the main steam line break and loss of
coolant accident have been revised. A brief summary of the
significant changes to the main steam line break analyses and the
radiological consequences of the main steam line break and loss of
coolant accident is presented below.
1. The limited fuel failure following a main steam line break
outside containment results in an increase in the calculated
radiological consequences both off-site and in the control room. To
limit the consequences of a main steam line break outside
containment, the
[[Page 66598]]
Technical Specification allowed steam generator tube leakage will be
reduced to 0.035 gpm [gallons per minute] per steam generator.
2. Credit will now be taken for iodine removal from the
containment atmosphere by the Containment Spray System (CSS). The
use of the CSS for iodine removal has not been previously approved
by the NRC.
3. The proposed increase to the allowable control room in-
leakage will provide additional operational flexibility to address
expected minor system degradation over time. The increase in the
allowable control room in-leakage will result in an increase in the
calculated dose to the Control Room Operators.
4. The addition of the dose consequences from containment sump
backleakage to the Refueling Water Storage Tank (RWST) has been
included in the off-site and control room loss of coolant accident
(LOCA) analyses increases the consequences of previously evaluated
accidents.
The containment sump backleakage into the RWST results in sump
water entering the RWST when the RWST is at its minimum level. The
RWST will become a radioactive source and contribute a shine dose to
the surrounding areas. The increase in dose rates onsite will not
prevent operators from remaining in the control room or from
accessing equipment needed to mitigate the accident.
All piping and valves associated with RWST backleakage are
located in a harsh radiation area. Backflow from the sump might
increase dose rates in the area where these components are located.
Additional dose contributions, where they occur, do not adversely
impact the environmental qualification of the vital equipment
located there. All vital equipment would continue to perform its
safety function.
5. Credit will be taken in the main steam line break analyses
for the recently installed cavitating venturis in the Auxiliary
Feedwater System. However, this will not change the amount of fuel
failure. Therefore, credit for this equipment will not impact the
radiological consequences of a main steam line break.
6. Credit will be taken for the Reactor Coolant System (RCS) low
flow reactor trip for the pre-scram inside containment main steam
line break analysis. This equipment will be qualified for the
expected containment environment following a main steam line break
inside containment and will be added to the Environmental
Qualification Master List.
7. Millstone Unit No. 1 design basis accidents, loss of coolant
and main steam line break, will no longer be evaluated for impact on
Millstone Unit No. 2 control room habitability. This credits the
decision to decommission Millstone Unit No. 1. [Footnote--B.D.
Kenyon letter to the NRC, ``Millstone Nuclear Power Station, Unit
No. 1 Certification of Permanent Cessation of Power Operations and
that Fuel Has Been Permanently Removed from the Reactor,'' dated
July 21, 1998.]
The revised main steam line break analyses and the revised
determinations of the radiological consequences of the main steam
line break and design basis LOCA analyses take credit for equipment
not previously assumed in the analyses, and for plant or equipment
operating restrictions not currently contained in the Technical
Specifications. The changes to the analyses will not adversely
affect the probability of an accident previously evaluated, but the
revised analyses results do indicate that the consequences of an
accident previously evaluated will increase. Specifically, the
following changes cause an increase in the consequences of an
accident previously evaluated.
1. The increase in allowable control room in-leakage from 100
SCFM [standard cubic feet per minute] to 130 SCFM when the Control
Room Emergency Ventilation System is operating in the recirculation/
filtration mode.
The dose to the Control Room Operators from a Millstone Unit No.
2 LOCA increased from 9.25 to 25.8 rem to the thyroid and from 0.205
to 2.29 rem to the skin. The dose to the whole body decreased. (Both
low wind speed and high wind speed release conditions were analyzed.
The low wind speed condition bounds the high wind speed condition.)
The dose to the Control Room Operators from a Millstone Unit No. 3
LOCA increased from 2.67 to 14 rem to the skin and from 0.209 to
1.484 rem to the whole body. The dose to the thyroid decreased. The
doses to the Control Room Operators from either a Millstone Unit No.
2 or Unit No. 3 LOCA remain below the GDC [General Design Criterion]
19 criteria of 30 rem thyroid, 5 rem whole body and 30 rem to the
skin.
The new calculated doses to the Millstone Unit No. 2 Control
Room Operators from a main steam line break outside containment are
29 rem thyroid, 0.03 rem whole body and 0.5 rem skin. The doses to
the Millstone Unit No. 2 Control Room Operators are below the GDC 19
criteria of 30 rem thyroid, 5 rem whole body, and 30 rem to the
skin. (Note: The dose to the Control Room Operators from a main
steam line break was not previously evaluated because fuel failure
was not predicted to occur.)
2. The limited fuel failure that is predicted in the revised
main steam line break analyses.
Previously, the radiological consequences of a main steam line
break were not determined and were not presented in the FSAR because
fuel failure was not predicted to occur. Because of the predicted
limited fuel failure for the main steam line break outside of
containment, the radiological consequences were analyzed. The
results to the Exclusion Area Boundary (EAB) are 4.8 rem thyroid and
0.06 rem whole body. The results to the Low Population Zone (LPZ)
are 2.3 rem thyroid and 0.02 rem whole body. To meet the dose
acceptance criteria to the Millstone Unit No. 2 Control Room
Operators, the maximum allowable Technical Specification primary to
secondary leak rate is being reduced to 0.035 gpm per steam
generator. The results to the Millstone Unit No. 2 Control Room
Operators are 29 rem thyroid, 0.03 rem whole body and 0.5 rem skin.
The main steam line break outside containment is the limiting
accident for the Millstone Unit No. 2 Control Room Operators.
However, the dose consequences of a main steam line break are less
than the 10CFR100 limits off-site of 300 rem thyroid and 25 rem
whole body, and the doses to the Millstone Unit No. 2 Control Room
Operators are below the GDC 19 criteria of 30 rem thyroid, 5 rem
whole body, and 30 rem to the skin.
3. Taking credit for the low RCS flow reactor trip for the pre-
scram inside containment main steam line break analysis.
Previous analyses did not credit the low RCS flow reactor trip
in a harsh environment. This credits the low flow trip in a manner
not previously reviewed by the NRC for Millstone Unit No. 2. Without
credit for this reactor trip, the predicted fuel failure for steam
line breaks inside containment would be higher.
4. Taking credit for the removal of radioactive iodine from the
containment atmosphere by containment spray.
Previous analyses did not rely on the spray function to reduce
iodine concentration in the post-accident atmosphere inside
containment. This adds a mitigation function to the CSS that has not
been previously reviewed by the NRC for Millstone Unit No. 2.
Without credit for the removal of iodine, the predicted dose
consequences following a LOCA would be higher.
5. The addition of sump backleakage to the RWST during a LOCA.
The resultant dose contribution to the LPZ from RWST backleakage
is 1.487 rem thyroid and 0.11 rem whole body. The total dose to the
LPZ from a design basis LOCA is 21.86 rem thyroid and 0.941 rem
whole body. The dose is well below the 10CFR100 limits of 300 rem
thyroid and 25 rem whole body. The dose to the EAB was not affected
because leakage into the RWST does not start until 25.45 hours post-
LOCA and the EAB is a 2-hour dose.
The resultant dose contribution to the Millstone Unit No. 2
Control Room Operators from RWST backleakage is 3.75 rem thyroid,
0.017 rem whole body and 0.296 to the skin. The total dose to the
Millstone Unit No. 2 Control Room Operators from the LOCA is 25.8
rem thyroid, 0.718 rem whole body and 2.29 rem to the skin. These
doses are below the GDC 19 limits of 30 rem thyroid and skin, and 5
rem whole body.
The analyses results meet the guidance contained in SRP
[Standard Review Plan] 15.1.5, SRP 15.6.5, and the limits of
10CFR100 and GDC 19. Therefore, there will be no significant
increase in the probability or consequences of an accident
previously evaluated.
Technical Specification Changes
Technical Specification Non-Technical Changes
The minor editorial and non-technical changes to correct
spelling (Technical Specification 3.3.2.1), modify the title of a
table column (Technical Specification 3.4.8), clarify the type of
measurement performed (Technical Specification 3.4.8), and establish
consistent terminology (Technical Specification 3.7.6.1) will not
result in any technical changes to the Millstone Unit No. 2
Technical Specifications. The proposed changes will have no adverse
effect on plant
[[Page 66599]]
operation. Therefore, there will be no significant increase in the
probability or consequences of an accident previously evaluated.
Technical Specification 3.4.6.2
The reduction in the maximum allowable value of primary to
secondary leakage per steam generator is consistent with the new
radiological assessment of the potential control room operator
exposure following a main steam line break outside of containment.
The wording change to SR [Surveillance Requirement] 4.4.6.2.1 will
clarify that the water inventory balance is used to verify
compliance with the identified and unidentified leakage limits.
Pressure boundary leakage would first show up as unidentified
leakage during performance of SR 4.4.6.2.1. Further investigation,
(plant walkdown) would be necessary to classify the unidentified
leakage as pressure boundary leakage. This is consistent with
established plant practices to detect pressure boundary leakage.
The addition of the new SR 4.4.6.2.2 will address the primary to
secondary leakage limit. The new SR will include an exception to
Technical Specification 4.0.4 that will allow the determination of
primary to secondary leakage to be deferred until after Mode 4 is
entered. Even though verification of compliance with the primary to
secondary limit will not be done prior to entering Mode 4, the limit
is still expected to be met.
The proposed changes will have no adverse effect on plant
operation. Therefore, there will be no significant increase in the
probability or consequences of an accident previously evaluated.
Technical Specification 3.4.8
The addition of the words ``of gross specific activity'' to the
Limiting Condition for Operation (LCO), Action Statements, and SR
will clarify what the E-Bar limit applies to. This is consistent
with the Technical Specification Definition (1.20) for E-Bar.
The addition of a footnote (*) to state the power history
requirements for the determination of E-Bar will ensure that the
necessary plant conditions are established prior to performing the
analysis. This will not affect the E-Bar LCO limit or the
requirement to perform the analysis. The proposed change is
consistent with NUREG--0212 and NUREG--1432.
The footnote will also specify that the provisions of
Specification 4.0.4 are not applicable. This will allow entry into
Mode 1, without determining the value of E-Bar, assuming that the
power history requirements will not be met until after Mode 1 is
entered. This will normally only apply following an extended
shutdown.
The Isotopic Analysis for Iodine (including I-131, I-133, and I-
135) sample requirement will be expanded to include the LCO
requirement for 100/E-Bar. This is consistent with the requirements
of Action Statement d. This change will expand the sampling
requirement for iodine. Minor wording changes will also be made to
be consistent with the proposed changes to the LCO wording.
The proposed changes will have no adverse effect on plant
operation. Therefore, there will be no significant increase in the
probability or consequences of an accident previously evaluated.
Technical Specification 3.6.2.1
The revised radiological assessment calculation for the design
basis accident credits iodine removal from the containment
atmosphere by the CSS. This will require a reduction in the allowed
outage time (AOT) of one containment spray train from seven days to
seventy two hours. This AOT is consistent with NUREG-0212 and NUREG-
1432. This will help ensure that plant equipment assumed in the
safety analyses will be available. This is a more restrictive change
which will have no adverse effect on plant operation. Therefore,
there will be no significant increase in the probability or
consequences of an accident previously evaluated.
Technical Specification 3.6.5.1
The value for the pressure drop across the combined HEPA [high-
efficiency particulate air] filters and charcoal adsorber banks
specified in SR 4.6.5.1.d.1 will be changed from a generic value
[less than or equal to] 6 inches water gauge) to a plant specific
value [less than or equal to] 2.6 inches water gauge). This is a
more restrictive change which will have no adverse effect on plant
operation. Therefore, there will be no significant increase in the
probability or consequences of an accident previously evaluated.
Technical Specification 3.7.6.1
The value for the pressure drop across the combined HEPA filters
and charcoal adsorber banks specified in SR 4.7.6.1.e.1 will be
changed from a generic value [less than or equal to] 6 inches water
gauge) to a plant specific value [less than or equal to] 3.4 inches
water gauge). This is a more restrictive change which will have no
adverse effect on plant operation.
SR 4.7.6.1.e.2 will be expanded to clarify that the test of the
capability of the Control Room Emergency Ventilation Trains to
switch to the recirculation mode is performed with the trains
initially operating in the normal mode and the smoke purge mode of
operation. This will not affect the requirement that the trains be
capable of switching to the recirculation mode.
The value of allowable control room air in-leakage specified in
SR 4.7.6.1.e.3 will be increased from 100 SCFM to 130 SCFM. This is
consistent with the recently revised control room radiological
analysis for the design basis accidents.
The proposed increase will provide additional operational
flexibility to address expected minor system degradation over time.
This increase is supported by the new analysis.
The proposed changes will have no adverse effect on plant
operation. Therefore, there will be no significant increase in the
probability or consequences of an accident previously evaluated.
Technical Specification 3.9.15
The value for the pressure drop across the combined HEPA filters
and charcoal adsorber banks specified in SR 4.9.15.d.1 will be
changed from a generic value [less than or equal to] 6 inches water
gauge) to a plant specific value [less than or equal to] 2.6 inches
water gauge). This is a more restrictive change which will have no
adverse effect on plant operation. Therefore, there will be no
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes have no adverse effect on how any of the
associated systems or components function to prevent or mitigate the
consequences of design basis accidents. Also, the proposed changes
have no adverse effect on any design basis accident previously
evaluated since the changes are consistent with the revised
analyses, and the appropriate acceptance criteria are met for the
revised analyses. Therefore, the license amendment request does not
impact the probability of an accident previously evaluated nor does
it involve a significant increase in the consequences of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes will not alter the plant configuration (no
new or different type of equipment will be installed) or require any
new or unusual operator actions. They do not alter the way any
structure, system, or component functions and do not alter the
manner in which the plant is operated. The proposed changes do not
introduce any new failure modes.
Also, the response of the plant and the operators following
these accidents is unaffected by the change. Therefore, the proposed
changes will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
Analyses Changes
The acceptance criteria for a main steam line break in the SRP
15.1.5 does not exclude the prediction of fuel failure. Instead, the
SRP requires that ``Any fuel damage calculated to occur must be of
sufficiently limited extent that the core will remain in place and
intact with no loss of core cooling.'' The limited fuel failure that
is now predicted in the revised main steam line break analyses meets
this acceptance criterion. In addition, the RCS low flow reactor
trip that is now being credited to function in a harsh environment
to limit fuel failure is already required to be operable by
Technical Specifications.
The revised dose consequences for the design basis accidents
assumes a control room in-leakage of 130 SCFM. In addition, iodine
removal by the CSS, which is already required to be operable by
Technical Specifications, is assumed. The acceptance criteria for
the dose consequences of the design basis accidents to the EAB, LPZ
and the control room personnel is met in the revised analyses.
Therefore, the revisions to the dose consequence analyses for the
design basis accidents do not involve a significant reduction in the
margin of safety.
[[Page 66600]]
Technical Specification Changes
The proposed changes will correct spelling and terminology
errors, reduce the maximum allowable primary to secondary leakage,
add a new surveillance requirement, modify surveillance requirements
for RCS specific activity, reduce the allowed outage time for a
containment spray train, reduce the allowed pressure drop across the
control room and enclosure building HEPA [high-efficiency
particulate air] filters, and increase the control room maximum
allowed in-leakage. These changes will have no adverse effect on
equipment important to safety. The equipment will continue to
function as assumed in the design basis accident analysis.
Therefore, there will be no significant reduction of the margin of
safety as defined in the Bases for the Technical Specifications
affected by these proposed changes.
The only adverse impact of the proposed changes is that the dose
consequences following an accident may increase. However, the
revised analyses show that the acceptance criteria for the accident
analyses are met. Therefore, based on the responses above, the
proposed changes are deemed safe.
The NRC has provided guidance concerning the application of
standards in 10CFR50.92 by providing certain examples (March 6,
1986, 51 FR 7751) of amendments that are considered not likely to
involve an SHC. The minor editorial and non-technical changes
proposed herein to correct reference, spelling, and terminology
errors are enveloped by example (i), a purely administrative change
to Technical Specifications. The changes proposed herein to add a
new surveillance requirement to verify primary to secondary leakage
and to reduce the allowable pressure drop across various ventilation
filters are enveloped by example (ii), a change that constitutes an
additional limitation, restriction, or control not presently
included in the Technical Specifications. All of the other changes
proposed herein are not enveloped by any specific example.
As described above, this License Amendment Request does not
impact the probability of an accident previously evaluated, does not
involve a significant increase in the consequences of an accident
previously evaluated, does not create the possibility of a new or
different kind of accident from any accident previously evaluated,
and does not result in a significant reduction in a margin of
safety. Therefore, NNECO has concluded that the proposed changes do
not involve an 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 Project Director: William M. Dean.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: October 22, 1998.
Description of amendment request: The licensee is proposing to
change Technical Specifications 3.3.2.1, ``Instrumentation--Engineered
Safety Feature Actuation System Instrumentation''; 3.4.9.3, ``Reactor
Coolant System [RCS]--Overpressure Protection Systems''; and 3.5.3,
``Emergency Core Cooling Systems--ECCS Subsystems--Tavg < 300 [degrees]
F.'' The proposed changes will allow Millstone Unit No. 2 to prevent an
automatic start of any high-pressure safety injection (HPSI) pump when
the shutdown cooling system (SDCS) is in operation (Mode 4 and below).
An inadvertent start of an HPSI pump could result in overpressurization
of the SDCS.
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:
In accordance with 10CFR50.92, Northeast Nuclear Energy Company
(NNECO) has reviewed the proposed changes and has concluded that
they do not involve a significant hazards consideration (SHC). The
basis for this conclusion is that the three criteria of
10CFR50.92(c) are not compromised. The proposed changes do not
involve an SHC because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to Technical Specifications 3.3.2.1 and
3.5.3 will no longer require the HPSI pump, required to be operable
in Mode 4, to start automatically on a Safety Injection Actuation
Signal (SIAS). (The automatic SIASs on low pressurizer pressure and
high containment pressure are not required to be operable in Mode 4.
However, the manual safety injection pushbuttons are required in
Mode 4). This will allow the operable HPSI pump control switch to be
placed in the pull-to-lock position without affecting the
operability of that pump. All HPSI pumps will be prevented from
automatically starting when the plant is in Mode 4, and the Shutdown
Cooling System (SDCS) is aligned to the RCS to prevent an
inadvertent start of a[n] HPSI pump which could overpressurize the
SDCS. These changes will not reduce the requirement for at least one
HPSI pump to be operable in Mode 4. The changes will require an
additional operator action to remove the operable HPSI pump breaker
control switch from the pull-to-lock position, in addition to
initiating safety injection by use of the manual pushbuttons, if
Safety Injection System actuation is needed in Mode 4. The
requirement to manually initiate a[n] HPSI pump, in addition to
manually initiating a[n] SIAS, does not involve complicated
equipment manipulations nor require extensive time for performing
the required operator actions. The HPSI pump control switches are
located in the Control Room on the same panels as the manual SIAS
pushbuttons. The additional step required to start a[n] HPSI pump
will not add any appreciable time for initiating HPSI flow while in
Mode 4. In addition, considering the lower probability of a
significant loss of coolant accident in Mode 4, and the slower plant
response to a loss of coolant accident in Mode 4, the time required
for the additional operator action will have no significant effect
on the consequences of the accident. Therefore, there will be no
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed change to Technical Specification 3.4.9.3,
Surveillance Requirement (SR) 4.4.9.3.3, will allow the use of the
new pull-to-lock feature of the HPSI pump control switches to
satisfy low temperature overpressure protection mass input
requirements. This will not affect either the LTOP [low-temperature
overpressure protection] HPSI pump mass input restrictions or the
level of control to ensure the HPSI pumps are not capable of
injecting into the RCS. The proposed changes will have no adverse
effect on plant operation. Therefore, there will be no significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed minor editorial and non-technical changes to add
amendment numbers to Page 3/4 3-12 and to revise the wording of SRs
4.4.9.3.2 and 4.4.9.3.3 will not result in any technical changes to
the Millstone Unit No. 2 Technical Specifications. The proposed
changes will have no adverse effect on plant operation. Therefore,
there will be no significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to the Bases reflect the proposed changes
to the applicable Technical Specifications. The proposed changes
will have no adverse effect on plant operation. Therefore, there
will be no 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 proposed changes will allow the use of the HPSI pump breaker
control switch
[[Page 66601]]
pull-to-lock feature. Operation of the HPSI pump in Mode 4 will
change since the operator will have to start the HPSI pump, in
addition to manually initiating safety injection. However, HPSI pump
operation is not an accident initiator. Therefore, the proposed
changes will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed Technical Specification changes will no longer
require the HPSI pump, required to be operable in Mode 4, to start
automatically on a[n] SIAS, will allow the use of the new pull-to-
lock feature of the HPSI pump control switches to satisfy low
temperature overpressure protection mass input requirements, and
will make minor editorial and non-technical changes. These changes
will have no adverse effect on equipment important to safety. The
equipment will continue to function as assumed in the design basis
accident analysis. Therefore, there will be no significant reduction
in the margin of safety as defined in the Bases for the Technical
Specifications affected by these proposed changes.
The only adverse impact of the proposed changes is that an
additional operator action will be necessary to initiate HPSI flow
in Mode 4, if needed. However, considering the lower probability of
a significant loss of coolant accident in Mode 4, and the slower
plant response to a loss of coolant accident in Mode 4, the time
required for the additional operator action will have no significant
effect on the consequences of the accident. Therefore, based on the
responses above, the proposed changes are deemed safe.
The NRC has provided guidance concerning the application of
standards in 10CFR50.92 by providing certain examples (March 6,
1986, 51 FR 7751) of amendments that are considered not likely to
involve an SHC. The minor editorial and non-technical changes
proposed herein to add page amendment numbers and clarify wording
are enveloped by example (i), a purely administrative change to
Technical Specifications. All of the other changes proposed herein
are not enveloped by any specific example.
As described above, this License Amendment Request does not
impact the probability of an accident previously evaluated, does not
involve a significant increase in the consequences of an accident
previously evaluated, does not create the possibility of a new or
different kind of accident from any accident previously evaluated,
and does not result in a significant reduction in a margin of
safety. Therefore, NNECO has concluded that the proposed changes do
not involve an 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 Project Director: William M. Dean.
PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating
Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: October 30, 1998.
Description of amendment request: Limerick Generating Station
(LGS), Units 1 and 2, Technical Specifications (TS) Surveillance
Requirements 4.8.4.3.b.1, 4.8.4.3.b.2, and 4.8.4.3.b.3 list the
Overvoltage (OV), Undervoltage (UV), and Underfrequency (UF) values for
the protective instrumentation for the RPS electric power monitoring
channels. The proposed changes correct a discrepancy between the
General Electric Nuclear Engineering (GENE) Design Specification for
Power Supply Monitoring Relays and the existing TS Allowable Values
(AVs). The changes will revise the OV, US, and UF values from 132VAC,
109VAC, and 57Hz to 127.6VAC, 110.7VAC, and 57.05Hz respectively.
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 Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed Tech Spec changes to section 4.8.4.3.b for the
Overvoltage (OV), Undervoltage (UV), and Underfrequency (UF) relays
are more conservative than the existing TS values. This change
provides more protection for the associated RPS components, thus
decreasing the probability of a failure in RPS. The associated Non-
Conformance Report and calculation provide assurance that the OV/UV/
UF settings are acceptable since the calculated values assure that
the RPS components will operate within their ratings. There are no
physical changes to the associated protective relays by the TS
change; thus, original design basis redundancy and separation is
maintained. There is no change in the interface of the RPS and its
power supplies.
The safety function of the RPS is to initiate a reactor scram in
order to protect the primary fission products barrier, the reactor
fuel. The proposed TS Change to impose more conservative Allowable
Values for the OV, UV, and UF relays will provide additional
assurance that the RPS will operate within equipment voltage and
frequency ratings, and will not be damaged by power system
anomalies. This change will not affect the scram function of RPS;
thus, the consequences of any design basis events will not be
affected.
Therefore, the proposed TS changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS Allowable Values changes will not result in any
physical changes to the RPS Electric Power Monitoring System.
Existing setpoints will not be changed, only the TS Allowable Values
are being modified to be more conservative.
The system redundancy and independence are not changed, and no
new failure modes are introduced.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
Currently, there are no TS bases for the existing RPS Electric
Power Monitoring System OV, UV, and UF allowable values. Specific
analytical limits for system voltage and frequency are not defined
in the Safety Analysis Report, nor discussed in any design basis
Allowed Outage Time or accident evaluation.
Investigation into the licensing basis has identified nominal
values of +/-10% of 120 VAC and -5% of 60 HZ for the Allowable
Values. These values are included in NUREG 0123, from which LGS's
TSs were developed. NUREG 0123 also provides no bases for these
values.
The proposed changes in the TS Allowable Values is based on a
revision to the calculation for RPS Breaker Panel--RPS / UPS
[uninterruptible power supply] System Bus Relay Settings. This
revision determines the new allowable values based on the design
ratings of RPS components, and factors in instrument inaccuracies
and margin. These changes will also provide bases for the associated
TS section. The proposed changes bring TSs into agreement with plant
design specifications.
Therefore, the proposed TS 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.
[[Page 66602]]
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, PA 19464.
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: Robert A. Capra.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of amendment request: October 22, 1998.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 4.8.2.1.b.3 to increase the minimum
battery electrolyte temperature limit from 60 deg.F to 72 deg.F. This
change resolves a discrepancy in the electrolyte temperature assumed in
the Class 1-E battery sizing calculations versus the limit specified in
the TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed TS change does not involve any physical changes to
plant structures, systems or components (SSC). The Class-1E
batteries will continue to function as designed. The Class-1E
battery system is designed to mitigate the consequences of an
accident, and therefore, can not contribute to the initiation of any
accident. The proposed TS surveillance testing and monitoring
requirements will continue to ensure that the Class-1E batteries are
capable of performing their required safety functions. In addition,
this proposed TS change will not increase the probability of
occurrence of a malfunction of any plant equipment important to
safety, since the manner i[n] which the Class-1E battery system is
operated is not affected by these proposed changes. The proposed
changes merely establish TS surveillance acceptance criteria that
more appropriately reflect the actual plant design. Therefore, the
proposed TS changes would not result in an increase of the
consequences of an accident previously evaluated.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
(2) The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed TS changes do not involve any physical changes to
the design of plant systems, structures or components. The design
and operation of the Class-1E battery system is not changed from
that currently described in the [Updated Final Safety Analysis
Report] UFSAR, only the allocation of battery capacity design margin
is affected by the increased TS minimum battery electrolyte
temperature limit. The Class-1E battery system will continue to
function as designed to mitigate the consequences of an accident.
Implementing new TS surveillance acceptance criteria that more
appropriately reflect the actual plant design does not permit plant
operation in a configuration that would create a different type of
malfunction to the Class-1E batteries than any previously evaluated.
In addition, the proposed TS changes do not alter the conclusions
described in the UFSAR regarding the safety related functions of the
Class-1E batteries or their support systems.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
(3) The proposed change does not involve a significant reduction
in a margin of safety.
The proposed TS change involves the implementation of new TS
surveillance acceptance criteria that more appropriately reflect the
actual plant design. The new TS minimum battery electrolyte
temperature limit enables the Class-1E battery capacity margin to be
allocated in a manner which conforms to Hope Creek's current
licensing basis. The ability of the Class-1E batteries to
independently supply their required loads for four hours without
support from battery chargers is not affected by these proposed
changes. The safety-related Class-1E support systems will ensure
that the proposed TS minimum electrolyte temperature limit is met.
Therefore, the proposed TS change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: Robert A. Capra.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of amendment request: October 15, 1998, as supplemented by
letter dated November 11, 1998.
Description of amendment request: The proposed amendments would
change the Vogtle Electric Generating Plant, Unit 1 and Unit 2 Facility
Operating Licenses to delete or modify certain license conditions,
which have become obsolete or inappropriate. In addition, the Technical
Specifications would be reconstituted to reflect revised word
processing. No change in technical requirements would be involved;
however, the font would be changed to Arial 11 point; page numbers
would be revised to a limiting condition for operation specific
numbering scheme; and intentional blank pages would be deleted.
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 either remove or modify provisions in the
VEGP [Vogtle Electric Generating Plant] Unit 1 and [Unit] 2
Operating Licenses that have been completed or are otherwise
obsolete. Each proposed change is summarized below:
Certain Surveillance Requirements (SRs) that were either added
or modified at the time of Improved Technical Specifications (ITS)
implementation were listed in the Operating Licenses with a schedule
for performance. With the exception of Unit 2 SR 3.8.1.20, all SRs
are deleted from the Operating Licenses, because they have since
been performed according to schedule, and will henceforth be
performed in accordance with the Technical Specifications.
A condition concerning changes to the Unit 1 initial test
program is deleted due to the completion of the program.
A condition related to FEMA [Federal Emergency Management
Agency] procedures and the emergency plan is deleted from the Unit 1
license due to the obsolescence of the condition.
Conditions requiring the submission of Unit 1 reports concerning
the steam generator tube rupture analysis, the reactor vessel level
instrumentation system, the safety parameter display system, the
detailed control room design review, and the zinc coating of the
diesel fuel storage tanks are deleted due to completion of the
required activities.
A condition requiring modification of the Unit 1 ventilation
exhaust of the alternate radwaste facility is deleted due to
completion of the required activity.
An exemption related to the seismic adequacy of the Unit 1 spent
fuel racks is deleted because the required actions are completed and
the exemption has been determined to be no longer in effect.
A condition in both the Unit 1 and Unit 2 licenses containing
reporting requirements for other license conditions is revised due
to ambiguities between the requirements in the license condition and
those published in NRC regulations.
A schedular exemption for the Unit 2 decommissioning funding
report is deleted
[[Page 66603]]
because the report was submitted as required and the exemption is no
longer in effect.
The Technical Specifications and associated Bases have been
converted from WordPerfect for DOS version 5.1 to
Microsoft Word 97. There were no changes to technical
requirements. The only visible changes to the document are as
follows: (1) the font was changed to Arial 11 point; [(2)] page
numbers were revised to an LCO [limiting condition for operation]
specific numbering scheme; and [(3)] intentionally blank pages were
deleted.
The proposed changes discussed above are strictly
administrative/editorial and do not affect the operation or function
of any plant system, component, or structure. Therefore, the
proposed changes do not increase the probability of occurrence or
the consequences of a previously evaluated accident.
2. The proposed changes do not create the possibility of a new
and different type of accident from any previously evaluated.
The proposed administrative/editorial changes do not alter the
operation of any plant system or equipment and do not introduce a
new mode of operation. Each requirement contained in the license
conditions proposed for deletion has either been completed or is
obsolete. Since these parts of the license are no longer applicable,
deletion of these items does not provide the potential for an
accident to be created. The conversion of the Technical
Specifications from one word processing format to another did not
involve any changes to technical requirements. Thus, the proposed
changes cannot create a new accident initiating mechanism, and do
not create the possibility of a new and different type of accident
from any previously evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The license conditions proposed for deletion are obsolete and
each requirement has been completed. The conversion of the Technical
Specifications from one word processing format to another did not
involve any changes to technical requirements. Since the proposed
changes are strictly administrative/editorial and do not involve any
physical or procedural changes to the plant, the margin of safety,
as defined in the bases for any Technical Specification is not
affected by the proposed 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: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia.
NRC Project Director: Herbert N. Berkow.
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: November 16, 1996 (TS 98-06).
Brief description of amendments: The proposed amendments would
change the Sequoyah Nuclear Plant Technical Specifications (TSs) by
revising the emergency diesel generator (EDG) surveillance requirements
(SRs) to add a note that allows the SR to be performed in Modes 1, 2, 3
or 4, if the associated components are already out-of-service for
testing or maintenance and to remove the SR that verifies certain
lockout features prevent EDG starting.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the Tennessee Valley
Authority (TVA), the licensee, has provided its analysis of the issue
of no significant hazards consideration, which is presented below:
TVA has concluded that operation of SQN Units 1 and 2, in
accordance with the proposed change to the TSs, does not involve a
significant hazards consideration. TVA's conclusion is based on its
evaluation, in accordance with 10 CFR 50.91(a)(1), of the three
standards set forth in 10 CFR 50.92(c).
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The probability of occurrence or the consequences for an
accident or malfunction of equipment is not increased by this
request. The proposal does not alter the way any structure, system
or component functions, does not modify the manner in which the
plant is operated, and does not alter equipment out-of-service time.
This request does not degrade the ability of the D/G [emergency
diesel generator] or equipment downstream of the load sequencers to
perform their intended function. Deleting the surveillance of a
nonsafety-related equipment protection function from TS likewise
does not change the probability or consequences of analyzed accident
scenarios. Dose consequences remain unchanged by this request.
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 FSAR [Final Safety Analysis
Report] is not created; nor is the possibility for an accident or
malfunction of a different type. The proposal does not alter the way
any structure, system or component functions and does not modify the
manner in which the plant is operated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The margin of safety has not been reduced since the test
methodologies are not being changed and LCO [Limiting Condition for
Operation] allowed outage times are not being changed. Deleting the
surveillance of a nonsafety-related equipment protection function
from TS likewise does not reduce the margin of safety. The results
of accident analysis remain unchanged by this request.
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 Project Director: Frederick J. Hebdon.
The Cleveland Electric Illuminating Company, Centerior Service Company,
Duquesne Light Company, Ohio Edison Company, Pennsylvania Power
Company, Toledo Edison Company, Docket No. 50-440, Perry Nuclear Power
Plant, Unit 1, Lake County, Ohio
Date of amendment request: October 27, 1998.
Description of amendment request: The proposed amendment would
modify the existing Minimum Critical Power Ratio (MCPR) Safety Limit
contained in Technical Specification 2.1.1.2. The change would apply
additional conservatism by modifying the MCPR Safety Limit values, as
calculated by General Electric, by maintaining the limit of 1.09 for
two recirculation loop operation and by increasing the limit from 1.10
to 1.11 for single loop operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
There is no change to any plant equipment. Per USAR Section
4.2.1, the fuel system design bases are provided in General Electric
Standard Application for Reactor Fuel (GESTAR II). The Minimum
Critical Power Ratio (MCPR) Safety Limit protects the fuel in
accordance with the design basis. The MCPR Safety Limit calculations
limit the bundle power to ensure the critical power ratio remains
unchanged. Therefore, there is not an increase in the probability of
transition boiling. The basis of the MCPR Safety Limit calculation
remains the same,
[[Page 66604]]
ensuring that greater than 99.9% of all fuel rods in the core avoid
transition boiling if the limit is not violated. Therefore, there is
no increase in the probability of the occurrence of a previously
analyzed accident.
The fundamental sequences of accidents and transients have not
been altered. The MCPR Operating Limits are selected such that
potentially limiting plant transients and accidents prevent the MCPR
from decreasing below the MCPR Safety Limit anytime during the
transient. Therefore, there is no impact on any of the limiting USAR
Appendix 15B transients. The radiological consequences are the same
as previously stated in the USAR, and as approved in the NRC Safety
Evaluation for GESTAR II. Therefore, the consequences of an accident
do not increase over previous evaluations in the USAR.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The MCPR Safety Limit values are designed to ensure that fuel
damage from transition boiling does not occur in at least 99.9% of
the fuel rods in the core as a result of the limiting postulated
accident. The values are calculated in accordance with GESTAR II and
the fuel vendor's interim implementing procedures, which incorporate
cycle-specific parameters.
The GESTAR II analysis has been accepted by the NRC as
comprehensive for ensuring that fuel designs will perform within
acceptable bounds. The MCPR Safety Limit ensures that the fuel is
protected in accordance with the design basis. The function,
location, operation, and handling of the fuel remain unchanged. In
addition, the initiating sequence of events has not changed.
Therefore, no new or different kind of accident is created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The MCPR Safety Limit values do not alter the design or function
of any plant system, including the fuel. The new MCPR Safety Limit
values were calculated using NRC-approved methods described in
GESTAR II and the fuel vendor's interim implementing procedures,
which incorporate cycle-specific parameters. The MCPR Safety Limit
values are consistent with GESTAR II, the NRC Safety Evaluation of
GESTAR II, the NRC Safety Evaluation Report for the Perry Nuclear
Power Plant and its Supplements for USAR Sections 4.4.1 and
15.0.3.3.1, and the Technical Specification Bases (Section 2.1.1.2)
for the MCPR Safety Limit. This change incorporates a cycle-specific
MCPR Safety Limit, as opposed to relying on the generic limit.
Therefore, the implementation of the proposed change to the MCPR
Safety Limit does not involve a 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: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Stuart A. Richards.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: October 27, 1998 (supersedes the April
12, 1996, amendment request). This notice supersedes the staff's
proposed no significant hazards consideration determination evaluation
for the requested changes that was published on May 8, 1996 (61 FR
20858).
Description of amendment request: The proposed amendment
application would change the technical specifications (TS) for the
reactor coolant system and associated Bases to allow the installation
of electrosleeves in the Callaway steam generators for two fuel cycles.
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 electrosleeve configuration has been designed and analyzed
in accordance with the requirements of the ASME [American Society of
Mechanical Engineers] Code. The applied stresses and fatigue usage
for the sleeve are bounded by the limits established in the ASME
Code. ASME Code minimum material property values are used for the
structural and plugging limit analysis. Mechanical testing has shown
that the structural strength of nickel electrosleeves under normal,
upset and faulted conditions provides margin to the acceptance
limits. These acceptance limits bound the most limiting (3 times
normal operating pressure differential) burst margin recommended by
RG [Regulatory Guide] 1.121. Leakage testing for \5/8\'', \7/8\'',
\11/16\'' and \3/4\'' tube sleeves has demonstrated that no
unacceptable levels of primary to secondary leakage are expected
during any plant condition.
The sleeve nominal wall thickness (used for developing the
depth-based plugging limit for the sleeve) is determined using the
guidance of Regulatory Guide 1.121 and the pressure stress equation
of Section III of the ASME Code. The limiting requirement of
Regulatory Guide 1.121, which applies to part throughwall
degradation, is that the minimum acceptable wall must maintain a
factor of safety of three against tube failure under normal
operating (design) conditions. A bounding set of design and
transient loading input conditions was used for the minimum wall
thickness evaluation in the generic evaluation. Evaluation of the
minimum acceptable wall thickness for normal, upset and postulated
accident condition loading per the ASME Code indicates these
conditions are bounded by the design condition requirement minimum
wall thickness.
A bounding tube wall degradation growth rate per cycle and a NDE
[Non-Destructive Examination] uncertainty has been assumed for
determining the sleeve TS plugging limit. The sleeve wall
degradation extent is determined by NDE. The degradation which would
require plugging sleeved tubes is developed using the guidance of RG
1.121 and is defined in BAW-10219P, to be 20% throughwall for any
service induced degradation.
The consequences of failure of the sleeve are bounded by the
current steam generator tube rupture analysis included in the
Callaway FSAR [Final Safety Analysis Report]. Due to the slight
reduction in diameter caused by the sleeve wall thickness, primary
coolant release rates would be slightly less than assumed for the
steam generator tube rupture analysis (depending on the break
location), and therefore, would result in lower total primary fluid
mass release to the secondary system.
A risk assessment for installation of Electrosleeves at Callaway
Plant was performed for a two-cycle operating period. The results of
this evaluation determined that sufficient margins against
postulated tube rupture during bounding accident conditions exist
for all types of degradation of the Electrosleeve material. The
calculated probability of burst for a hypothetical population of
10,000 axial flaws, 100% throughwall of the parent tube and 0.40''
long, is 4.4 x 10-11 at the end of the second operating cycle. The
probability of burst for postulated circumferential flaws and pits
is determined to be essentially zero.
The proposed change does not adversely impact any other
previously evaluated design basis accident or the results of LOCA
[Loss of Coolant Accident] and non-LOCA accident analyses for the
current technical specification minimum reactor coolant system flow
rate. The results of the analyses and testing demonstrate that the
electrosleeve is an acceptable means of maintaining tube integrity.
Furthermore, per Regulatory Guide 1.83 recommendations, the sleeved
tube can be monitored through periodic inspections with present NDE
techniques. These measures demonstrate that installation of sleeves
spanning degraded areas of the tube will restore the tube to a
condition consistent with its original design basis.
Conformance of the electrosleeve design with the applicable
sections of the ASME Code and results of the leakage and mechanical
tests, support the conclusion that installation of electrosleeves
will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
[[Page 66605]]
Electrosleeving does not represent a potential to adversely
affect any plant component. Stress and fatigue analysis of the
repair has shown that the ASME Code and Regulatory Guide 1.121
criteria are not exceeded. Implementation of electrosleeving
maintains overall tube bundle structural and leakage integrity at a
level consistent to that of the originally supplied tubing during
all plant conditions. Leak and mechanical testing of electrosleeves
support the conclusions of the calculations that each sleeve retains
both structural and leakage integrity during all conditions.
Sleeving of tubes does not provide a mechanism resulting in an
accident outside of the area affected by the sleeves. Any accident
as a result of potential tube or sleeve degradation in the repaired
portion of the tube is bounded by the existing tube rupture accident
analysis.
Implementation of sleeving will reduce the potential for primary
to secondary leakage during a postulated steam line break while not
significantly impacting available primary coolant flow area in the
event of a LOCA. By effectively isolating degraded areas of the tube
through repair, the potential for steam line break leakage is
reduced. These degraded intersections now are returned to a
condition consistent with the Design Basis. While the installation
of a sleeve reduces primary coolant flow, the reduction is far below
that caused by plugging. Therefore, far greater primary coolant flow
area is maintained through sleeving versus plugging.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The electrosleeve repair of degraded steam generator tubes has
been shown by analysis to restore the integrity of the tube bundle
consistent with its original design basis condition, i.e., tube/
sleeve operational and faulted condition stresses are bounded by the
ASME Code requirements and the repaired tubes are leaktight. The
safety factors used in the design of sleeves for the repair of
degraded tubes are consistent with the safety factors in the ASME
Code used in steam generator design. The portions of the installed
sleeve assembly which represent the reactor coolant pressure
boundary can be monitored for the initiation and progression of
sleeve/tube wall degradation, thus satisfying the requirements of
Regulatory Guide 1.83. The portion of the tube bridged by the sleeve
is effectively removed from the pressure boundary, and the sleeve
then forms the new pressure boundary. The areas of the sleeved tube
assembly which require inspection are defined in BAW-10219P.
In addition, since the installed sleeve represents a portion of
the pressure boundary, a baseline inspection of these areas is
required prior to operation with sleeves installed. The effect of
sleeving on the design transients and accident analyses has been
reviewed based on the installation of sleeves up to the level of
steam generator tube plugging coincident with the minimum reactor
flow rate and the Callaway Safety Analysis.
Provisional requirements cited in other NRC Safety Evaluation
Reports addressing the implementation of sleeving have required the
reduction of the individual steam generator normal operation primary
to secondary leakage limit from 500 to 150 gpd [gallons per day].
Consistent with these evaluations, Union Electric will reduce the
per steam generator leak rate of 500 gpd in TS 3.4.6.2.c to 150 gpd.
The establishment of this leakage limit at 150 gpd provides
additional safety margin. [The staff notes that this leakage limit
has been incorporated into the Callaway Technical Specifications via
license amendment #119 dated October 1, 1996.]
Finally, Union Electric will reduce the tube plugging limit from
48% through wall to 40% through wall to be consistent with NUREG-
1431. The establishment of the plugging limit at 40% through wall
provides additional safety margin. [The staff notes that this
plugging limit has been incorporated into the Callaway Technical
Specifications via license amendment #119 dated October 1, 1996.]
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: University of Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
NRC Project Director: William H. Bateman.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont
Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: November 3, 1998.
Description of amendment request: The licensee proposes to make
administrative changes to the Technical Specifications to correct
errors, add consistency within the Technical Specifications, and make
nomenclature changes to support and enhance usability of the Technical
Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated, because:
The proposed changes are purely administrative in nature and
have no effect on plant hardware, plant design, safety limit
setting, or plant system operation and therefore do not modify or
add any initiating parameters that would significantly increase the
probability or consequences of an accident previously evaluated.
No new modes of operation are introduced by the proposed changes
such that adverse consequences would result. Accordingly, the
consequences of previously analyzed accidents are not affected by
this proposed license amendment.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated, because:
These changes do not affect the operation of any systems or
components, nor do they involve any potential initiating events that
would create any new or different kind of accident. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated for the
Vermont Yankee Nuclear Power Station.
3. Involve a significant reduction in a margin of safety,
because:
These proposed changes do not affect any equipment involved in
potential initiating events or safety limits. Therefore, it is
concluded that the proposed changes do not involve a significant
reduction in a margin of safety.
Administrative changes, as such, do not constitute any
significant hazards considerations.
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: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Project Director: Cecil O. Thomas.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: November 10, 1998.
Description of amendment request: The proposed changes to North
Anna Power Station (NAPS), Units 1 and 2, Technical Specification (TS)
3.4.4 will clarify the operability requirements for the pressurizer
heaters and eliminate a potential verbatim compliance issue associated
with the pressurizer heaters and emergency power supply. The verbatim
compliance issue was created when the Emergency Diesel Generator
allowed outage time was changed from 72 hours to 14 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the
[[Page 66606]]
licensee has provided its analysis of the issue of no significant
hazards consideration, which is presented below:
Virginia Electric and Power Company has reviewed the
requirements of 10 CFR 50.92 as they relate to the proposed changes
for the North Anna Units 1 and 2 and determined that a significant
hazards consideration is not involved. The proposed changes will
revise the LCO [limiting condition for operation] 3.4.4 to require
that the pressurizer have two groups of pressurizer heaters operable
with a capacity of greater than or equal to 125 kW and capable of
being powered from its associated emergency bus. The Action
Statement will also be revised to focus on heater operability. The
following is provided to support this conclusion.
(a) Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The pressurizer heaters are not an initiator of any accident
previously evaluated. As a result, the probability of any accident
previously evaluated is not increased. The pressurizer heaters
remain operable as assumed in the accident analysis to mitigate the
consequences of any accident. Therefore, the proposed changes to
clarify the operability requirements do not significantly increase
the probability of occurrence or the consequences of any previously
analyzed accident.
(b) Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed Technical Specifications changes do not involve any
physical alteration of the plant or changes in methods governing
normal plant operation. Operation of and the design of the
pressurizer heaters and the associated power supplies are not
changed by the proposed changes. The proposed changes do not impose
any new or eliminate any existing requirements. Therefore, it is
concluded that no new or different kind of accident or malfunction
from any previously evaluated has been created.
(c) Does the change involve a significant reduction in a margin
of safety?
The proposed Technical Specifications changes will not reduce
the margin of safety since the change has no effect on any safety
analyses assumptions. The pressurizer heaters remain operable as
assumed in the safety analysis to mitigate the consequences of any
accident previously analyzed. The proposed changes only clarify the
operability requirements for the pressurizer heaters and associated
emergency power supplies. Therefore, the proposed changes do not
result in a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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 Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams,
Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia
23219.
NRC Project Director: Herbert N. Berkow.
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Florida Power and Light Company, et al., Docket Nos. 50-335, and 50-
389, St. Lucie Plant, Unit Nos. 1, and 2, St. Lucie County, Florida
Date of amendment request: October 29, 1998.
Description of amendment request: Technical Specification changes
(TS) relating to the implementation and automatic removal of certain
reactor protection system trip bypasses to ensure that the meaning of
explicit terms used in the TSs are consistent with the intent of the
stated requirements.
Date of publication of individual notice in the Federal Register:
November 5, 1998 (63 FR 59809).
Expiration date of individual notice: November 19, 1998.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see: (1) The
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, 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 No. 50-318, Calvert Cliffs
Nuclear Power Plant, Unit No. 2, Calvert County, Maryland
Date of application for amendment: July 20, 1998.
Brief description of amendment: The amendment implements a
modification that constitutes an unreviewed safety question as
described in 10 CFR 50.59. The modification involves replacing the
service water heat exchangers with new plate and frame heat exchangers
having an increased thermal performance capability. The planned
modification is similar to the one completed on Unit 1. In addition, by
a separate letter dated July 20, 1998, the licensee submitted a request
to obtain approval for a temporary one time cooling lineup needed to
support emergency diesel generator operability for the installation of
the Unit 2 service water heat exchanger replacement, which is currently
being reviewed by the NRC
[[Page 66607]]
staff. Therefore, since the implementation of the proposed service
water heat exchanger modification is dependent on the staff's issuance
of the one time Technical Specification (TS) change regarding
installation of the modification, this modification should not be
implemented prior to the issuance of the one-time TS change for
installing the modification.
Date of issuance: November 5, 1998.
Effective date: This license amendment is effective as of the date
of its issuance to be implemented after the staff's issuance of the
one-time TS change regarding the installation of the service water heat
exchanger modification.
Amendment No.: 203.
Facility Operating License No. DPR-69: Amendment revised the
Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: August 12, 1998 (63 FR
43201).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 5, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: June 26, 1998.
Brief description of amendment: The amendment modifies various
Technical Specification pages to correct typographical errors, remove
inadvertent replication of information, and updates various Bases
sections.
Date of issuance: November 10, 1998.
Effective date: November 10, 1998.
Amendment No: 178.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 23, 1998 (63
FR 50933).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 10, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
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: March 6, 1998, as supplemented
September 11, 1998. The September 11, 1998, supplemental letter
contained clarifying information only, and did not change the no
significant hazards consideration determination.
Brief description of amendment: The amendment revises Technical
Specification 3.9.2 relating to the use of Post-Accident Monitoring
Source Range neutron flux detectors as a compensatory measure in the
event that one of the two required BF3 neutron flux detectors becomes
inoperable during Mode 6 operations (refueling).
Date of issuance: November 12, 1998.
Effective date: November 12, 1998.
Amendment No: 180.
Facility Operating License No. DPR-23: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 3, 1998 (63 FR
30262).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 12, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: September 17, 1998, as
supplemented October 15, 1998.
Brief description of amendments: The amendments revised the Updated
Final Safety Analysis Report to perform a Keowee Emergency Power
Engineered Safeguards Functional Test during the 1998 Unit 3 refueling
outage at Oconee.
Date of Issuance: November 12, 1998.
Effective date: As of the date of issuance to be implemented during
the 1998 Unit 3 refueling outage.
Amendment Nos.: Unit 1--233; Unit 2--233; Unit 3--232.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:
Amendments revised the Updated Final Safety Analysis Report.
Date of initial notice in Federal Register: September 30, 1998 (63
FR 52304).
The October 15, 1998, letter provided clarifying information that
did not change the scope of the September 17, 1998, 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 November 12, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear
Station, Dauphin County, Pennsylvania
Date of application for amendment: December 2, 1996.
Brief description of amendment: This amendment would revise audit
frequency requirements and relocate them from the Technical
Specifications to the Quality Assurance Plan.
Date of issuance: November 12, 1998.
Effective date: This amendment is effective immediately to be
implemented written 60 days.
Amendment No.: 52.
Facility Operating License No. DPR-73: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 30, 1997 (62 FR
40850).
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
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: May 11, 1998, as supplemented
by letter dated October 9, 1998.
Brief description of amendments: The amendments modify the
technical specifications (TS) for San Onofre Nuclear Generating Station
Unit Nos. 2 and 3 to implement 10 CFR Part 50 Appendix J, Option B for
performance-based reactor containment leakage testing.
Date of issuance: November 6, 1998.
Effective date: November 6, 1998, to be implemented within 30 days
from the date of issuance.
Amendment Nos.: Unit 2 -144; Unit 3 -135.
Facility Operating License No. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
[[Page 66608]]
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48265).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 6, 1998.
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: July 6, 1998.
Brief description of amendments: Relocates the description of the
reactor coolant system design features in Technical Specification 5.4
to the Updated Final Safety Analysis Report, which already contains the
information.
Date of issuance: November 18, 1998.
Effective date: November 18, 1998, to be implemented within 30
days.
Amendment Nos.: Unit 1--Amendment No. 98; Unit 2--Amendment No. 85.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48266).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 18, 1998.
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: July 6, 1998 , as supplemented on
October 28, 1998.
Brief description of amendments: Relocate the Technical
Specification 3/4.3.3.3 requirements for Seismic Instrumentation to the
Technical Requirements Manual.
Date of issuance: November 18, 1998.
Effective date: November 18, 1998, to be implemented within 30
days.
Amendment Nos.: Unit 1--Amendment No. 99; Unit 2--Amendment No. 86.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48267).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 18, 1998.
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: July 6, 1998, as supplemented on October
28, 1998.
Brief description of amendments: Relocates the Technical
Specification 3/4.7.13 requirements for the Area Temperature Monitoring
System to the Technical Requirements Manual.
Date of issuance: November 18, 1998.
Effective date: November 18, 1998, to be implemented within 30
days.
Amendment Nos.: Unit 1--Amendment No. 100; Unit 2--Amendment No.
87.
Facility Operating License Nos. NPF-76 and NPF-80: The amendment
revises the Technical Specifications.
Date of initial notice in Federal Register: September 9, 1998 (63
FR 48267). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 18, 1998.
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.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee.
Date of application for amendments: February 13, 1998 (TS 97-07).
Brief description of amendments: The amendments incorporate new
main steam isolation valve (MSIV) requirements that are consistent with
NUREG-1431, the Westinghouse Standard Technical Specifications (TS),
including testing requirements for the MSIVs that ensure the valves
close on an automatic actuation signal.
Date of issuance: November 17, 1998.
Effective date: As of the date of issuance to be implemented no
later than 45 days after issuance.
Amendment Nos.: 236 and 226.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19980).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 17, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendments: June 26, 1998 (TS 98-02).
Brief description of amendments: The amendments change the
Technical Specifications and their Bases to lower the specific activity
limit for the primary coolant system from 1.0 microcurie/gram dose
equivalent iodine-131 to 0.35 microcurie/gram, as provided for in NRC
Generic Letter 95-05, ``Voltage-Based Repair Criteria for Westinghouse
Steam Generator Tubes Affected by Outside Diameter Stress Corrosion
Cracking.'' This change allows a proportional increase in main steam
line break induced primary-to-secondary leakage when implementing the
alternate steam generator tube repair criteria, which the NRC has
already approved for Sequoyah Nuclear Plant, Units 1 and 2.
Date of issuance: November 17, 1998.
Effective date: As of the date of issuance to be implemented no
later than 45 days after issuance.
Amendment Nos.: 237 and 227.
Facility Operating License Nos. DPR-77 and DPR-79: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: July 15, 1998 (63 FR
38205).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 17, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
[[Page 66609]]
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, (WBN) Rhea County, Tennessee
Date of application for amendment: August 5, 1998 (TS 98-008).
Brief description of amendment: This amendment is in response to
your application dated August 5, 1998. The amendment revises the WBN
Technical Specifications (TS) and associated TS Bases to allow up to 4
hours to make the residual heat removal suction relief valve available
as a cold overpressure mitigation system relief path.
Date of issuance: November 10, 1998.
Effective date: November 10, 1998.
Amendment No.: 14.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 23, 1998 (63
FR 50940).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 10, 1998.
No significant hazards consideration comments received: None.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: July 10, 1996 (TXX-96405), as
supplemented by letters dated October 1, 1996 (TXX-96475), and July 1,
1998 (TXX-98159).
Brief description of amendments: The amendment would take credit
for the addition of train oriented Fan Coil Units for each UPS and
Distribution Room and would provide redundancy to the existing Air
Conditioning (A/C) Units (TS 3/4.7.11 and its associated bases).
Date of Issuance: Date of issuance: November 18, 1998.
Effective date: November 18, 1998, to be implemented within 30
days.
Amendment Nos.: Unit 1--Amendment No. 61; Unit 2--Amendment No. 47.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 12, 1997 (62
FR 6579).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 18, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: May 7, 1998.
Brief description of amendment: This amendment revises Technical
Specification 5.4, ``Fuel Storage,'' to increase the allowable mass of
uranium-235 (U235) per axial centimeter for fuel storage.
The requested change will allow the use of new Siemens Power
Corporation heavy fuel assembly designs.
Date of Issuance: November 12, 1998.
Effective date: November 12, 1998.
Amendment No.: 141.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 17, 1998 (63 FR
33111).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 12, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
Dated at Rockville, Maryland, this 24th day of November 1998.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director Division of Reactor Projects--III/IV Office of Nuclear
Reactor Regulation.
[FR Doc. 98-31931 Filed 12-1-98; 8:45 am]
BILLING CODE 7590-01-P