[Federal Register Volume 64, Number 56 (Wednesday, March 24, 1999)]
[Notices]
[Pages 14278-14292]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-7032]
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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 March 1, 1999, through March 12, 1999. The
last biweekly notice was published on March 10, 1999 (64 FR 11958).
Notice of Consideration of Issuance of Amendments to Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administration Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the NRC Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By April 23, 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
[[Page 14279]]
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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of amendments request: December 16, 1998.
Description of amendments request: The proposed amendment would
revise Technical Specification (TS) 3.8.1, ``AC Sources--Operating,''
and TS 3.3.7, ``Diesel Generator (DG)--Loss of Voltage Start (LOVS).''
The proposed amendment will (1) change Condition G of TS 3.8.1 to
ensure that the appropriate actions will be taken to prevent double
sequencing of safety-related loads, and (2) change TS 3.3.7 to ensure
that the setpoint allowable values for the degraded voltage and the
loss of voltage relays reflect the required function of the relays.
Basis for proposed no significant hazards consideration determination:
As required by 10 CFR 50.91(a), the licensee has provided its analysis
of the issue of no significant hazards consideration, which is
presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendment will change Condition G of Technical
Specification 3.8.1. These changes will ensure that the appropriate
actions will be taken to prevent double sequencing of safety-related
loads. This change is required to assure the capability of the
offsite circuits ``to effect a safe shutdown and to mitigate the
effects of an accident'' in accordance with Regulatory Guide 1.93.
The proposed amendment will also change the setpoint allowable
values for the degraded voltage and the loss of voltage relays in
Technical Specification Surveillance Requirement (SR) 3.3.7.3. The
proposed changes do not involve any physical changes to plant
equipment. The actions required by the TS amendment will identify
when an offsite circuit does not meet its required capability and
provides actions to restore the required capability. The proposed
changes are intended to identify and correct the conditions (voltage
and/or loading) required to prevent the possibility of a double
sequencing event. Therefore, this change ensures that power will be
supplied to the ESF [engineered safety feature] loads following a
loss of offsite power event described in UFSAR [Updated Final Safety
Analysis Report] 15.2.6.1. For other events discussed in the UFSAR,
the electrical distribution system is an event mitigator. This
change will ensure that the electrical distribution system will
continue to meet this requirement. The proposed changes will not
effect the function of the DG loss of voltage start as required by
the design basis and safety analysis. Therefore, the proposed change
does not involve a significant increase in the probability of an
accident previously evaluated.
The proposed changes do not involve any physical changes to
plant equipment. The proposed changes ensure that appropriate
controls are in place to prevent a double sequencing event. The
proposed changes consider the factors in preventing a double
sequencing event such as pretrip voltage, load, number of units on
line, and number of transmission lines in service. These are factors
which could affect post trip voltage. The actions associated with
this change will identify and mitigate the condition where an
offsite circuit does not meet its required capability and, as such,
do not result in new or revised accident sequences. The proposed
changes will not effect the function of the DG loss of voltage start
as required by the design basis and safety analysis. Therefore, the
proposed change does not involve a significant increase in the
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 amendment will change Condition G of Technical
Specification 3.8.1. These changes will ensure that the appropriate
actions will be taken to prevent double sequencing of safety-related
loads. This change is required to assure the capability of the
offsite circuits ``to effect a safe shutdown and to mitigate the
effects of an accident'' in accordance with Regulatory
[[Page 14280]]
Guide 1.93. The proposed amendment will also clarify the setpoint
allowable values for the degraded voltage and the loss of voltage
relays in Technical Specification Surveillance Requirement (SR)
3.3.7.3. The proposed changes do not change the operation of any
system or equipment, nor do they create a new type of malfunction.
The proposed changes prevent double sequencing and do not create the
possibility of any other malfunction. The actions associated with
this change will identify and mitigate the condition where an
offsite circuit does not meet its required capability. The proposed
changes will not effect the function of the DG loss of voltage start
as required by the design basis and safety analysis. Therefore, the
proposed changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed amendment will change Condition G of Technical
Specification 3.8.1. These changes will ensure that the appropriate
actions will be taken to prevent double sequencing of safety-related
loads. This change is required to assure the capability of the
offsite circuits ``to effect a safe shutdown and to mitigate the
effects of an accident'' in accordance with Regulatory Guide 1.93.
The proposed amendment will also change the setpoint allowable
values for the degraded voltage and the loss of voltage relays in
Technical Specification Surveillance Requirement (SR) 3.3.7.3. The
proposed changes ensure that the units will be in conformance with
GDC 17, Electric Power Systems (basis for TS 3.8.1). The required
actions of the proposed change will ensure that the single failure
analyses and safety analysis are maintained. The actions associated
with this change will identify and mitigate the condition where an
offsite circuit does not meet its required capability. The proposed
changes ensure that the bases for the current TS are maintained. The
proposed changes will not effect the function of the DG loss of
voltage start as required by the design basis and safety analysis.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
NRC Project Director: William H. Bateman.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: February 26, 1999.
Description of amendment request: The proposed amendment would
revise the Table Notations for Technical Specification (TS) Table 3.3-
4, ``Engineered Safety Features Actuation System Instrumentation Trip
Setpoints.'' Specifically, the time constants used in the lead-lag
controller for Steam Line Pressure--Low (Table item 1.e.) are
t1 greater than or equal to 50 seconds and t2
greater than or equal to 5 seconds. The proposed amendment would revise
t2 to less than or equal to 5 seconds. Also, the time
constant used in the rate-lag controller for Negative Steam Line
Pressure Rate--High (Table item 4.e.) is less than or equal to 50
seconds. The proposed amendment would revise this time constant to
greater than or equal to 50 seconds.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Correcting the time constants will ensure conservative
calibration of the Engineered Safety Feature Actuation System
instrumentation. The proposed amendment will not introduce any new
equipment or require existing equipment to function different from
that previously evaluated in the Final Safety Analysis Report (FSAR)
or TS. Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Correcting the time constants will ensure conservative
calibration of the Engineered Safety Feature Actuation System
instrumentation. The proposed amendment will not introduce any new
equipment or require existing equipment to function different from
that previously evaluated in the Final Safety Analysis Report (FSAR)
or TS. The proposed amendment will not create any new accident
scenarios, because the change does not introduce any new single
failures, adverse equipment or material interactions, or release
paths. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
Correcting the time constants will ensure conservative
calibration of the Engineered Safety Feature Actuation System
instrumentation. Therefore, the proposed change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Cecil Thomas.
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of application for amendment: March 26, 1997.
Brief description of amendment: The amendment revises modifies
Technical Specification sections 3.6 and 4.5 by removing the list of
containment isolation valves in accordance with Generic Letter 91-08,
``Removal of Components Lists from Technical Specifications,'' dated
May 6, 1991, and by revising requirements related to containment
pressure and containment temperature. Additionally, several editorial
changes are made to emulate the format and content of NUREG-1432,
``Standard Technical Specifications, Combustion Engineering Plants.''
Date of issuance: February 22, 1999.
Effective date: February 22, 1999.
Amendment No.: 184.
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 17, 1997 (62
FR 66136)
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 22, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
[[Page 14281]]
Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: September 3, 1997.
Description of amendment request: The proposed amendment includes
the following changes to the station technical specification (TS):
(a) TS Action Statement 3.14a is replaced by a revised condition
description for TS Action Statement 3.17.1.6 in the instrumentation
systems section. Also, the maximum control room temperature at which a
shutdown must be initiated is revised from 120 deg.F [degrees
Fahrenheit] to 90 deg.F, and a time limit for reaching the hot
shutdown condition is specified;
(b) TS 3.14b is replaced with two limiting conditions for operation
(LCOs), 3.14.1 and 3.14.2, addressing, respectively, the filtration and
cooling functions of the CRHVAC [control room heating, ventilation, and
air conditioning] system. These proposed LCOs emulate the standard TS
(NUREG 1432) for control room ventilation;
(c) TS Table 4.2.3 surveillance requirement (SR) number 3,
verification of control room temperature, is moved to SR Table 4.17.1,
for the reactor protection system (RPS); and
(d) other administrative changes.
The licensee classified each change as either administrative or
more restrictive. An administrative change is editorial in nature,
involves only movement of requirements within the TS without affecting
their technical content, or clarifies existing TS requirements. A more
restrictive change adds new requirements, or revises existing
requirements resulting in more conservative or additional operational
restrictions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes to TS 3.14a and TS 3.14b constitute either
new, or more restrictive requirements that provide additional assurance
that equipment conforms to the plant design basis and will operate
reliably when called upon. These changes represent additional
restrictions on plant operation that enhance safety and are consistent
with the standard TS. The proposed change to TS Table 4.2.3 of moving
SR item number 3 to TS Table 4.17.1, and other administrative changes
are editorial in nature or involve the reorganization or reformatting
of TS requirements without affecting technical content or operational
restrictions. The proposed changes do not result in any substantive
change in operating requirements or the intent of these requirements,
and are consistent with the Commission's regulations. Therefore, these
changes cannot involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
The proposed changes to TS 3.14a and TS 3.14b constitute either
new, or more restrictive requirements that provide additional assurance
that equipment conforms to the plant design basis and will operate
reliably when called upon. These changes represent additional
restrictions on plant operation that enhance safety and are consistent
with the standard TS. The proposed change to TS Table 4.2.3 of moving
SR number 3 to TS Table 4.17.1, and other administrative changes are
editorial in nature or involve the reorganization or reformatting of TS
requirements without affecting technical content or operational
restrictions. The proposed changes do not result in any substantive
change in operating requirements or the intent of these requirements,
and are consistent with the Commission's regulations. Therefore, these
changes cannot create the possibility of a new or different kind of
accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
The proposed changes to TS 3.14a and TS 3.14b constitute either
new, or more restrictive requirements that provide additional assurance
that equipment conforms to the plant design basis and will operate
reliably when called upon. These changes represent additional
restrictions on plant operation that enhance safety and are consistent
with the standard TS. The proposed change to TS Table 4.2.3 of moving
SR number 3 to TS Table 4.17.1, and other administrative changes are
editorial in nature or involve the reorganization or reformatting of TS
requirements without affecting technical content or operational
restrictions. The proposed changes do not result in any substantive
change in operating requirements or the intent of these requirements,
and are consistent with the Commission's regulations. Therefore, these
changes cannot involve a significant reduction in a margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423-3698.
Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Project Director: Cynthia A. Carpenter.
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: March 1, 1999.
Description of amendment request: The proposed amendments would
revise Oconee Nuclear Station, Units 1, 2, and 3 Improved Technical
Specification (ITS) 3.3.8 to only require two channels for the reactor
coolant system hot leg temperature function. The current TSs require
two channels per loop. This requirement was incorrectly specified
during the ITS conversion.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated:
The proposed change modifies ITS Table 3.3.8-1 to only require
two channels for RCS [Reactor Coolant System] Hot Leg Temperature
Function. These instruments provide indication only and are not
considered as initiators of any analyzed event. The proposed change
does not involve a physical alteration of the plant. No new or
different equipment is being installed, and no installed equipment
is being operated in a new or different manner. No set points for
parameters which initiate protective or mitigative action are being
changed. Therefore, the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any kind of accident previously evaluated:
The proposed change does not involve a physical alteration of
the plant. No new or different equipment is being installed, and no
[[Page 14282]]
installed equipment is being operated in a new or different manner.
No set points for parameters which initiate protective or mitigative
action are being changed. As a result, no new failure modes are
being introduced. Therefore, this proposed amendment will not create
the possibility of any new or different kind of accident.
3. Involve a significant reduction in a margin of safety.
The margin of safety for PAM [post accident monitoring]
instrumentation is based on the availability and capability of the
instrumentation to provide the required operator information. The
proposed change maintains requirements within the safety analyses
and licensing basis and has no effect on the availability and
capability of the PAM function. Therefore, the 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: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Attorney for licensee: Ann W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC.
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: March 1, 1999.
Description of amendment request: The proposed amendments to
Improved Technical Specification (ITS) 3.9, ``Refueling Operations,''
Subsection 3.9.3, ``Containment Penetrations,'' Limiting Condition for
Operation 3.9.3.b would add a Note to state that the emergency air lock
door is not required to be closed when it is sealed with a temporary
cover plate. The temporary cover plate contains penetrations that are
used for such refueling outage services as cables, pneumatic tubing,
and hoses.
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:
This proposed change has been evaluated against the standards in
10 CFR 50.92 and has been determined to involve no significant
hazards, in 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?
The proposed change allows the use of a temporary cover plate as
a seal for the emergency air lock during refueling operations in
lieu of an air lock door. Duke [Duke Energy Corporation] analyses
for Oconee Nuclear Station (ONS) does not credit containment
closure. Therefore, use of the temporary cover plate does not affect
offsite doses, which were previously calculated to be well within 10
CFR 100 limits. As such, the proposed change does not involve a
significant increase in the probability or consequences of an
accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The fuel handling accident inside containment analyses discussed
in the Updated Final Safety Analysis Report section 15.11 bound the
proposed change. No new or different type of accident will occur
because of the temporary cover plate placement.
3. Involve a significant reduction in a margin of safety.
Placing the temporary cover plate in the emergency air lock will
still meet the intent of containment closure. The building pressure
does not increase during a fuel handling accident and fission
products will be contained. The fuel handling accident inside
containment analyses does not credit containment closure for
reducing offsite dose. As such, the proposed change does not involve
a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC.
NRC Project Director: Herbert N. Berkow.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: February 24, 1999.
Description of amendment request: The proposed amendments would
change Technical Specification (TS)
3/4.7.4 to remove the restriction to monitor the Ultimate Heat Sink
(UHS) temperature only in the Intake Cooling Water (ICW) bay and prior
to the ICW pumps. This change would permit the option of monitoring the
UHS temperature after the ICW pumps but prior to the component cooling
water heat exchangers, which is considered to be equivalent to
temperature monitoring before the ICW pumps.
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 method of monitoring the Ultimate Heat Sink temperature is
not considered in, and has no effect on, the probability of any type
of accident initiating sequence. The proposed changes will permit
other means of monitoring the Ultimate Heat Sink that have been
evaluated to be equivalent to the current method permitted. As the
monitoring will continue to be performed by equal means, the
consequences of any accident previously evaluated will not be
affected.
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 change will permit other means of monitoring the
Ultimate Heat Sink temperature, which will be equal to the methods
currently employed. The continued monitoring of this variable by
equivalent means cannot create the possibility of a new or different
type of accident.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The Ultimate Heat Sink temperature is an input assumption used
in the accident analysis and in evaluation of component design. This
temperature limit is not being altered by this change, only the
permissible means of monitoring this variable. As any new methods
employed are expected to be equivalent to those currently used, no
reduction in any margin of safety will result.
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: Florida International
University, University Park, Miami, Florida 33199.
Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,
P.O. Box 14000, Juno Beach, Florida 33408-0420.
NRC Project Director: Cecil O. Thomas.
[[Page 14283]]
GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear
Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: February 2, 1999.
Description of amendment request: The proposed amendment revises
the Technical Specifications (TS) to expand the scope of systems and
test requirements considered under TS 4.5.4 ``Engineered Safeguards
Feature (ESF) Systems Leakage,'' and increases the maximum allowable
leakage for those portions of the ESF system outside containment. The
proposed amendment also includes revised the Bases for TS 3.15.3,
``Auxiliary and Fuel Handling Building Air Treatment System,'' to
clarify system design requirements and accident analysis
considerations.
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. No
physical modifications which would change structures, systems or
components are proposed by this TSCR [technical specification change
request] for surveillance changes in Technical Specification 4.5.4
and its Bases. The proposed increase in the ESF Systems leakage rate
acceptance limit has no effect on the performance of ESF systems
during a DBA [design basis accident]. The proposed changes are
supported by a revised MHA [maximum hypothetical accident] dose
calculation using updated X/Q values and calculation assumptions.
The MHA dose consequence analysis yields dose results that are below
the 10 CFR 100 guidelines for both the EAB [exclusion area boundary]
and LPZ [low population zone]. The calculated Control Room
Habitability Evaluation does not exceed the permissible annual
occupational exposure limit of 50 Rem to the thyroid as specified in
10 CFR 20.1201(a)(ii). In addition, the potential thyroid exposure
can be mitigated by the availability of self-contained breathing
apparatus and potassium iodide. Therefore, the changes would not
involve a significant increase in the consequences of accidents
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 previously evaluated. This TSCR does not
involve any physical modifications that would affect structures,
systems, or components, nor does it involve any changes in plant
operation.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. This TSCR does not involve changes to the Technical
Specification defined Safety Limits, Limiting Conditions for
Operation, and does not involve any change to safety system
setpoints for operation. Therefore, the proposed changes do not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (Regional Depository) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Elinor G. Adensam.
Northern States Power Company, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: February 12, 1999.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TS) to (1) allow reactor vessel
hydrostatic and leakage tests without maintaining primary containment
integrity, (2) establish a limit and a surveillance requirement on
reactor coolant activity when reactor coolant temperature is above
212 deg.F, the reactor is not critical, and primary containment has not
been established, and (3) correct a punctuation error.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes do not increase the probability of an
accident since reactor vessel hydrostatic and leakage tests would be
performed with the reactor vessel nearly water solid, at nominal
operating pressure, not critical and at low decay heat values which
minimizes the energy stored in the reactor vessel. Under this
proposed change a limit on reactor coolant activity is established
that provides adequate assurance that the consequences of a large
primary system break during reactor vessel hydrostatic and leakage
test conditions will be conservatively bounded by the consequences
of a postulated main steam line break outside of primary
containment. Low pressure emergency core cooling systems are
required to be operable during reactor vessel hydrostatic and
leakage test providing assurance that adequate core cooling can be
achieved to preclude fuel failures and subsequent increases in
reactor coolant activity in the event of a large primary system
break. The reduced stored energy in the reactor vessel and proposed
limit on reactor coolant activity ensures there is no increase in
the probability or consequences of an accident previously evaluated.
The proposed amendment will not create the possibility of a new
or different kind of accident from any accident previously analyzed.
The proposed changes do not introduce any new accident
initiators or failure mechanisms since the changes do not involve
any changes to the structures, systems, or components. They also do
not involve any change to the operation of systems, and alter
procedures only to the extent that 212 deg.F may be exceeded during
reactor vessel hydrostatic and leakage testing without maintaining
primary containment integrity. Without maintaining primary
containment integrity, a large primary system break during a reactor
vessel hydrostatic or leakage test would result in the same kind of
accident as would a main steam line break outside primary
containment during normal operation. Therefore, the proposed TS
change does not create the possibility of a new or different kind of
accident, from any accident previously evaluated.
The proposed amendment will not involve a significant reduction
in the margin of safety.
Since reactor vessel hydrostatic and leakage tests are performed
nearly water solid, at nominal operating pressure, not critical and
at low decay heat values, the stored energy in the reactor vessel
during testing will be low. Under these conditions, the potential
for failed fuel and a subsequent increase in coolant activity is
minimized. Therefore, the proposed Technical Specification change
does not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. In addition, correction to the punctuation error is strictly
a grammatical change and has no effect on the three standards of 10 CFR
50.92(c). Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and
[[Page 14284]]
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of amendment request: February 8, 1999.
Description of amendment request: The proposed amendments would
revise Technical Specification 4.5.3.2.b to allow the option of using
closed and disabled automatic valves to provide the necessary isolation
function when performing safety injection and charging pump testing in
Modes 4, 5, and 6 (hot shutdown, cold shutdown, and refueling) for low
temperature over pressurization protection.
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. Will not involve a significant increase in the probability or
consequences of an accident previously evaluated.
In Mode 4 with the RCS [reactor coolant system] coolant
temperature less than 312 deg.F or in Modes 5 and 6 there is a
potential risk of low temperature overpressurization. Mass additions
of coolant by the safety injection and charging pumps could cause
such an event to the extent that these pump flows exceed the ability
of a single over pressure protection relief valve to protect the
system. In order to eliminate this potentiality provisions are made
to allow a maximum of one pump to be in service with the other pumps
disabled except for testing. Further provisions are made to assure
that a pump being tested can not inject into the vessel. The
proposed change merely adds an alternate method of providing this
assurance in addition to that currently provided by closing the
manual discharge valves. The proposed change offers an equivalent
means of affording the required protection.
Based upon the above, the proposed change will not increase the
probability or consequences of an accident previously analyzed.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
The proposed changes do not require any change in the operation
of the plant. A minor configuration change is involved in that a[ ]
disabled automatic valve in the flow path will be used in lieu of
the manual valve to provide protection. Specifically, no new
hardware is being added to the plant as part of the proposed change,
no existing equipment is being modified, and no significant changes
in operations are being introduced. Therefore, these changes will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. Will not involve a significant reduction in a margin of
safety.
The proposed change will not alter any assumptions, initial
conditions, or results of any accident analyses. The proposed change
maintains the level of protection. The change will, therefore, 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: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Project Director: Elinor G. Adensam.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of amendment request: March 1, 1999.
Description of amendment request: The proposed amendment would
revise the Ginna Station Improved Technical Specifications battery cell
parameters limit for specific gravity Surveillance Requirement (SR)
3.8.6.3 and SR 3.8.6.6.
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 Ginna Station in accordance with the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated. The change is only
to correct an error in the determination of the minimum limiting
value for specific gravity of the station batteries. This does not
increase the probability of an accident previously evaluated since
the battery specific gravity is only a measure of the state of
charge of the battery and the batteries themselves are not an
accident initiator. The proposed minimum value for specific gravity,
based on the NUREG-1431 guidance, gives a higher assurance that the
battery has sufficient capacity. Therefore, the probability or
consequences of an accident previously evaluated is not
significantly increased.
(2) Operation of Ginna Station in accordance with the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated. The proposed change
does not involve a physical alteration of the plant (i.e. no new or
different type of equipment will be added) or changes in the methods
governing normal plant operation. The change only involves
implementing a more conservative minimum limiting value for the
battery cell parameter of specific gravity. Therefore, the
possibility for a new or different kind of accident from any
accident previously evaluated is not created.
(3) Operation of Ginna Station in accordance with the proposed
change does not involve a significant reduction in a margin of
safety. The proposed change only corrects an error in the
determination of the limiting value for specific gravity. The error
is being corrected by using a more conservative value as determined
by the guidance of NUREG-1431. Therefore, this 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: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400
L Street, NW., Washington, DC 20005.
NRC Project Director: S. Singh Bajwa.
South Carolina Electric & Gas Company (SCE&G), South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station,
Unit No. 1, Fairfield County, South Carolina
Date of amendment request: February 18, 1999.
Description of amendment request: The proposed amendment would
revise Virgil C. Summer Nuclear Station (VCSNS) Technical Specification
(TS) 3/4.4.9 Reactor Coolant System Pressure/Temperature Limits to
incorporate the new Pressure/Temperature (PT) Limits curves consistent
with reactor vessel specimen analysis results. Additionally, the
proposed amendment would revise the Pressure/Temperature Limits Bases
section to accurately reflect current industry standards and
regulations.
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 changes revise the Pressure/Temperature Limits
Curves to provide curves that reflect the results of the analysis
performed on reactor vessel surveillance
[[Page 14285]]
specimen W. This analysis was performed using NRC approved
methodology as documented in WCAP 14040-NP-A, dated January, 1996.
These curves provide the limits for operation of the Reactor Coolant
System during heat up, cool down, criticality, and hydrotesting. The
limits protect the reactor vessel from brittle fracture by
separating the region of acceptable operation from the region where
brittle fracture is postulated to occur. Failure of the reactor
vessel is not a VCSNS design basis accident, and, in general,
reactor vessel failure has a low probability of occurrence and is
not considered in the safety analysis.
Therefore, the change 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?
The proposed changes revise the Pressure/Temperature Limits
Curves, Section 3/4.4.9, to incorporate the results of the analysis
performed on reactor vessel specimen W. There are no plant design
changes or significant changes in any operating procedures. This
change adjusts the heat up and cool down curves to reflect the shift
in nil-ductility reference temperature of the reactor vessel as a
result of neutron embrittlement. Therefore, the change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does this change involve a significant reduction in margin of
safety? The proposed changes revise the Pressure/Temperature Limits
Curves, Section 3/4.4.9, to incorporate the results of the analysis
performed on reactor vessel specimen W. The new PT curves ensure
that the 10 CFR 50 Appendix G, requirements are not exceeded during
normal operation including Reactor Coolant System transients during
heat up, cool down, criticality, and hydrotesting. The new PT curves
were prepared, using approved NRC methodology, for a projected
reactor vessel neutron exposure of 32 EFPY [effective full power
years].
The new curves shift to more conservative operating limitations,
thus providing increased margin against non-ductile fractures. Since
administrative limits remain in place to ensure that 10 CFR 50
Appendix G limits are not challenged, the margin of safety described
in the TS Bases is not reduced by the proposed change. Therefore,
the 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: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180.
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218.
NRC Project Director: Herbert N. Berkow.
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 amendment requests: May 8, 1996, as supplemented by letter
dated January 13, 1999.
Description of amendment requests: The January 13, 1999,
supplemental letter added an additional change to the technical
specifications (TS) to incorporate an additional restriction to the
time required to close containment when reactor coolant system (RCS)
water level is reduced during a refueling outage. This additional
restriction adds a limitation that containment must be able to be
closed within the calculated time to boil, if it is less than the
current four hour requirement. The January 13, 1999, letter supplements
the staff's proposed no significant hazards consideration determination
evaluation that was published on September 11, 1996 (61 FR 47978).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The licensee's analysis of the issue of no significant
hazards consideration on the supplemental change is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Item 6 conservatively restricts the completion time to ensure
containment closure is achieved prior to the water in the cavity
boiling, in the event of a Loss of Shutdown Cooling. This
restriction is already a self imposed requirement at San Onofre
Units 2 and 3. Incorporating it in the Technical Specification only
serves to highlight the importance of this requirement.
This change captures all periods of time when the time to boil
following a Loss of Shutdown Cooling is less than 4 hours. Having
this requirement cannot initiate an accident. However, this
requirement reduces the consequences of a Loss of Shutdown Cooling
Accident when the time to boil is less than 4 hours.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Item 6 conservatively restricts the completion time to ensure
containment closure is achieved prior to the water in the cavity
boiling, in the event of a Loss of Shutdown Cooling. This
restriction is already a self imposed requirement at San Onofre
Units 2 and 3. Incorporating it in the Technical Specification only
serves to highlight the importance of this requirement.
This restriction cannot initiate an accident.
3. The proposed change does not involve a significant reduction
in a margin of safety.
Item 6 conservatively restricts the completion time to ensure
containment closure is achieved prior to the water in the cavity
boiling, in the event of a Loss of Shutdown Cooling. This
restriction is already a self imposed requirement at San Onofre
Units 2 and 3. Incorporating it in the Technical Specification only
serves to highlight the importance of this requirement.
This change increases the margin of safety provided by the
Technical Specification by specifying that the containment must be
closed within 4 hours or within the calculated time to boil,
whichever is less. This change revises the Technical Specification
to specifically recognize the importance of ensuring containment
closure is achieved prior to boiling in the reactor vessel, upon a
loss of shutdown cooling.
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 requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, P.O. Box 19557, Irvine, California 92713.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, P. O. Box 800, Rosemead, California 91770.
NRC Project Director: William H. Bateman.
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 amendment requests: December 22, 1998.
Description of amendment requests: The proposed amendment would
modify the technical specifications (TS) to add a reference to allow
use of Westinghouse laser-welded steam generator (SG) tube sleeving.
The proposed amendment also provides typographical and editorial
corrections.
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.
Steam generator tubes, tube plugging, and tube failures are
considered in the analysis of
[[Page 14286]]
accidents in the Updated Final Safety Analysis Report (UFSAR). The
steam generator tube rupture accident analysis considered the
failure of a steam generator tube. Also, inadvertent opening of a
steam generator dump valve (IOSGDV), loss of condenser vacuum
(LOCV), loss of coolant accidents (LOCAs), and feed water line break
(FWLB) accident analyses carry assumptions regarding steam generator
tube plugging. In each case, the addition of steam generator tube
sleeves to repair defective tubes will not change the probability or
consequences of any accident previously evaluated.
The sleeve configurations have been designed, analyzed, and
tested in accordance with the American Society of Mechanical
Engineers (ASME) code requirements, and mechanical testing has shown
that the sleeve and sleeve joints provide margin above acceptance
limits. Ultrasonic testing (UT) and eddy current testing (ECT) are
used to verify the adequacy of welds. Tests have demonstrated that
tube collapse will not occur due to postulated LOCA loadings.
The probability or consequences of any accident previously
evaluated is not increased because any leakage through the sleeve
assembly is fully bounded by the existing steam generator tube
rupture analysis included in the San Onofre Unit 2 and 3 Updated
Final Safety Analysis Report. Additionally, any reactor coolant flow
restriction from sleeving is addressed by a ratio of number of
sleeved tubes to be equal to a plugged tube.
Therefore, the proposed sleeving repair process will not involve
an increase in the probability or consequences of any previously
evaluated accident.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The sleeves are captured within the steam generator tubes by
hard rolling and welding and as such are not able to physically
affect other parts of the system. The failure of a sleeve is
identical to the failure of the parent tube which has been
previously analyzed.
The use of a sleeve to span the area of degradation of the steam
generator tube restores the structural and leakage integrity of the
tubing to meet the original design requirements. Structural analysis
of the sleeve assembly shows that the requirements of the ASME code
are met. Mechanical testing has demonstrated that margin exists
above the original tube design criteria. Any hypothetical accident
as a result of any degradation in a sleeved tube would be bounded by
the existing steam generator tube rupture accident analysis.
Therefore, operation of the facility in accordance with proposed
changes does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The use of sleeves to repair degraded steam generator tubing
will maintain the integrity of the tube bundle commensurate with the
ASME Code and draft Regulatory Guide (RG) 1.121 margin requirements
for original tubing. Sleeves are components which are part of the
reactor coolant pressure boundary and meet the requirements for
Class 1 components in Section III of the ASME Boiler and Pressure
Vessel Code. The primary to secondary pressure boundary will be
maintained to the same margins as the original tubes under normal
and postulated accident conditions. The safety margins used in the
verification of the strength of the sleeve assembly are consistent
with the safety factors in the ASME Boiler and Pressure Vessel Code
used in steam generator design. Further, a test program has been
conducted by Westinghouse which demonstrated the integrity of the
lower hard rolled joint design and its capability to withstand the
design loads.
Therefore, operation of the facility with the proposed changes
will not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, P.O. Box 19557, Irvine, California 92713.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, P. O. Box 800, Rosemead, California 91770.
NRC Project Director: William H. Bateman.
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 26, 1999 (TS 98-08).
Brief description of amendments: The proposed amendments would
change the Sequoyah (SQN) Technical Specifications (TS) by relocating
TS 3.7.6, ``Flood Protection Plan,'' and the associated bases to the
SQN Technical Requirements Manual (TRM). This change does not alter the
current requirements for implementation or surveillance testing of the
Flood Protection Plan and future revisions of this plan will require an
evaluation in accordance with 10 CFR 50.59.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), Tennessee Valley
Authority (TVA), the licensee, has provided its analysis of the issue
of no significant hazards consideration, which is presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed revision to the TS relocates the requirements for
SQN flood protection without changing the current requirements. This
administrative relocation of the requirements will not increase the
possibility of an accident.
The capability of the Flood Protection Plan will continue to
provide the same function. Changes to the relocated requirements
will be processed, in accordance with 10 CFR 50.59, to ensure the
Flood Protection Plan will be properly maintained. Therefore, the
proposed relocation of the flood protection requirements will not
increase the probability or consequences of an accident previously
evaluated.
The proposed amendment does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The SQN Flood Protection Plan is used to mitigate the effects of
a flooding event at SQN. This plan would not be the initiator of any
new or different kind of accident. The capability of the Flood
Protection Plan will continue to provide the same function. Changes
to the relocated requirements will be processed, in accordance with
10 CFR 50.59, to ensure the Flood Protection Plan will be properly
maintained. The proposed change does not alter the current functions
of SQN's Flood Protection Plan; therefore, this proposed change will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
B. The proposed amendment does not involve a significant
reduction in a margin of safety.
The requirements for SQN's flood protection are unchanged by the
proposed relocation of the requirements to the SQN TRM. The function
of the Flood Protection Plan and surveillance requirements to ensure
implementation of the plan remains unchanged. Any future changes to
these requirements will be evaluated, in accordance with 10 CFR
50.59, to ensure acceptability and NRC review as required.
Accordingly, the proposed change will not result in a reduction in a
margin of safety.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
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: Cecil O. Thomas.
[[Page 14287]]
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 26, 1999 (TS 99-02).
Brief description of amendments: The proposed amendments would
change the Sequoyah (SQN) Technical Specifications (TS) to provide for
consistency when exiting the action statements associated with the
Emergency Diesel Generators (D/Gs). The Tennessee Valley Authority
(TVA) inadvertently omitted revising Action Statements c, d, and e
associated with TS 3.8.1.1 in Revision 1 to TS Change 96-08, addressing
the D/G allowed outage time, submitted to the NRC staff on October 8,
1998.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), TVA has provided its
analysis of the issue of no significant hazards consideration, which is
presented below:
A. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
This proposed revision provides for consistency and removes
contradictions within the action statements associated with TS
3.8.1.1. Additionally, the proposed revision will not result in any
change in the design, maintenance or operation of the associated
plant equipment nor will it result in deviation from the actions
presently approved by the staff for SQN's response to the associated
LCOs [Limiting Conditions for Operation]. The deletion of the
defined portion of the requirements associated with the restoration
of offsite power sources in Action Statements c and d does not
result in any change to SQN's response to the stated condition since
this requirement remains unchanged in Action Statement a.
The deletion of the requirements associated with the restoration
of 4 diesel generator (D/G) sets within 72 hours from Action
Statements c and e provides for a consistent allowed outage time of
7 days for the loss of a D/G set as previously approved by the staff
in a safety evaluation issued on December 16, 1998. Therefore, the
proposed amendment does not involve an increase in the probability
or consequences of an accident previously evaluated.
B. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed change provides for consistency and removes
contradictions within the action statements associated with TS
3.8.1.1. Additionally, the proposed revision will not result in any
change in the design, maintenance or operation of the associated
plant equipment nor will it result in deviation from the actions
presently approved by the staff for SQN's response to the associated
LCOs. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
C. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change provides for consistency and removes
contradictions within the action statements associated with TS
3.8.1.1. Additionally, the proposed revision will not result in any
change in the design, maintenance or operation of the associated
plant equipment nor will it result in deviation from the actions
presently approved by the staff for SQN's response to the associated
LCOs. Therefore, the proposed amendment does not involve a reduction
in a margin of safety.
The NRC has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
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: Cecil O. Thomas.
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.
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone Nuclear
Power Station, Unit 1, New London County, Connecticut
Date of application of amendment: December 4, 1998, January 18, and
January 19, 1999.
Brief description of amendment: The proposed amendment would modify
the staffing and training requirements to allow the use of Certified
Fuel Handlers to meet plant staffing requirements.
Date of publication individual notice in Federal Register: December
29, 1998 (63 FR 71657).
Expiration date of individual notice: January 28, 1999.
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.
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.
[[Page 14288]]
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, as supplemented
December 4, 1998, and December 23, 1998.
Brief description of amendment: The amendment permits a one-time
change to the Technical Specification (TS) Bases for TS 3.8.2 for
Calvert Cliffs Nuclear Power Plant, Unit No. 2 and provides approval of
the licensee's analysis of unreviewed safety questions as described in
10 CFR 50.59. The change allows Baltimore Gas and Electric Company to
provide alternate cooling to the Unit 2 emergency diesel generators
(EDGs) during their replacement of the Unit 2 service water (SRW) heat
exchangers in the 1999 refueling outage since the normal SRW cooling
would be unavailable. The licensee proposes to provide the 2A EDG with
cooling water from the Unit 1 SRW system and to provide the 2B EDG with
cooling water from an independent external cooling system during the
replacement work.
Date of issuance: March 8, 1999.
Effective date: As of the date of its issuance to be implemented
during the Calvert Cliffs Unit No. 2 spring 1999 refueling outage.
Amendment No.: 205.
Facility Operating License No. DPR-69: Amendment revised the
Technical Specifications Bases.
Date of initial notice in Federal Register: August 26, 1998 (63 FR
45523) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: October 9, 1998.
Brief description of amendment: The amendment revised Section 6.0
to Technical Specifications to change the membership of the Nuclear
Facility Safety Committee and corrected other typographical errors.
Date of issuance: March 8, 1999.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 199.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 16, 1998 (63
FR 69337).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: April 9, 1998 (NRC-98-0071).
Brief description of amendment: The amendment revises Technical
Specification (TS) 3.7.1.2, ``Emergency Equipment Cooling Water
System,'' Action a, and TS 3.8.1.1, ``A.C. Sources--Operating,'' Action
c, to be consistent with the actions required for inoperable oxygen
monitoring instrumentation in TS 3.3.7.5, ``Accident Monitoring
Instrumentation.'' The existing ``**'' footnote to TS 3.7.1.2, Action
a, is modified and a ``*'' footnote is added to TS 3.8.1.1, Action c.
Date of issuance: March 3, 1999.
Effective date: March 3, 1999, with full implementation within 30
days.
Amendment No.: 132.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 23, 1998 (63
FR 50937).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: September 23, 1998.
Brief description of amendment: The amendment changes Division III
battery specific gravity acceptance criteria outlined in River Bend
Station (RBS) Technical Specifications (TS). The change is required as
a result of Division III battery system modifications scheduled to be
implemented during refueling outage RF-8, beginning April 3, 1999.
During this time, the current Division III battery will be replaced
with a new battery having a greater capacity rating. The new battery
has a nominal specific gravity of 1.215 at 77 deg.F in contrast to the
existing Division III battery supplied with a nominal specific gravity
of 1.210 at 77 deg.F. Since TS Section 3.8.6, Table 3.8.6-1 values for
specific gravity are based upon the manufacturer's nominal specific
gravity, these values were updated to reflect the changes.
Date of issuance: March 3, 1999.
Effective date: The license amendment is effective upon the date of
issuance and shall be implemented within 90 days.
Amendment No.: 103.
Facility Operating License No. NPF-47: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64111).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: August 31, 1998.
Brief description of amendment: This amendment revised Technical
Specification Surveillance Requirement 3.6.1.3.4 to permit removal of
the inclined fuel transfer system primary containment blind flange
while primary containment integrity is required.
Date of issuance: February 24, 1999.
Effective date: February 24, 1999.
Amendment No.: 100.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56260).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 24, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
[[Page 14289]]
FirstEnergy Nuclear Operating Company, Docket No. 50-440 Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: July 13, 1998, and as
supplemented by submittal dated November 23, 1998.
Brief description of amendment: This amendment revised Technical
Specification 3.4.4,'' Safety/Relief Valves (SRVs),'' by increasing the
present plus or minus 1% tolerance on the safety mode lift setpoint for
the safety relief valves to plus or minus 3%.
Date of issuance: March 3, 1999.
Effective date: March 3, 1999.
Amendment No.: 101.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 12, 1998 (63 FR
43214).
The supplemental information contained clarifying information and
did not change the initial no significant hazards consideration
determination and did not expand the scope of the original application.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, OH 44081.
Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit
1, DeWitt County, Illinois
Date of application for amendment: October 5, 1998.
Brief description of amendment: The amendment allows deferral of
the next scheduled local leak rate test for valve 1MC-042 until the
seventh refueling outage.
Date of issuance: March 8, 1999.
Effective date: March 8, 1999, and shall be implemented within 45
days.
Amendment No.: 121.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 23, 1998 (63 FR
56949).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 8, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, IL 61727.
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone Nuclear
Power Station, Unit 1, New London County, Connecticut
Date of application for amendment: December 4, 1998, and January 18
and 19, 1999.
Brief description of amendment: The proposed amendment would modify
the staffing and training requirements to allow the use of Certified
Fuel Handlers to meet plant staffing requirements.
Date of issuance: March 5, 1999.
Effective date: As of the date of issuance to be implemented within
45 days from the date of issuance.
Amendment No.: 104.
Facility Operating License No. DPR-21: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 29, 1998 (63
FR 71657).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 5, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of applications for amendment: August 12, 1998, as
supplemented by letter dated October 30, 1998; and application dated
September 28, 1998, as supplemented by letters dated January 7 and 20,
1999.
Brief description of amendment: The amendment allows implementation
of a revised main steamline break analysis and revised control room
habitability analyses.
Date of issuance: March 10, 1999.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 228.
Facility Operating License No. DPR-65: Amendment revised the
Technical Specifications and authorized changes to the Final Safety
Analysis Report.
Date of initial notice in Federal Register: October 7, 1998 (63 FR
53951) and December 2, 1998 (63 FR 66597).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 10, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of application for amendment: December 10, 1998, as
supplemented February 19, 1999.
Brief description of amendment: The amendment allows the licensee
to implement changes to the Final Safety Analysis Report (FSAR)
regarding a revised method for ensuring boron precipitation can be
prevented (post-loss-of-coolant accident).
Date of issuance: March 10, 1999.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 229.
Facility Operating License No. DPR-65: Amendment authorizes changes
to the Final Safety Analysis Report.
Date of initial notice in Federal Register: January 13, 1999 (64 FR
2249).
The February 19, 1999, supplemental letter provided additional
information that did not change the staff's proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 10, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: June 10, 1998, as supplemented
October 30, 1998.
Brief description of amendment: The amendment revises the Millstone
Unit 3 licensing basis associated with post-accident mitigation
activities, vital area access travel routes, and the associated action
completion times. Northeast Nuclear Energy Company determined that the
Final Safety Analysis Report (FSAR) description of post-accident vital
area routing was out of date
[[Page 14290]]
because the radiological control area boundary fence created an access
problem on the designated routes to the hydrogen recombiner and fuel
building. The revised licensing basis will be incorporated into the
FSAR and will revise the routes to accommodate the fence location and
allow for the time to unlock gates.
Date of issuance: March 1, 1999.
Effective date: As of the date of issuance to be implemented within
60 days from the date of issuance.
Amendment No.: 166.
Facility Operating License No. NPF-49: Amendment authorized
revision to the FSAR.
Date of initial notice in Federal Register: July 15, 1998 (63 FR
38202).
The October 30, 1998, letter provided clarifying information that
did not change the scope of the June 10, 1998, application, and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: December 4, 1998.
Brief description of amendment: The amendment eliminates the need
to cycle the plant and its components through a shutdown-startup cycle
by allowing the next snubber surveillance interval to be deferred until
the end of refueling outage 6 or September 10, 1999, whichever date is
earlier.
Date of issuance: March 3, 1999.
Effective date: As of the date of issuance to be implemented within
30 days from the date of issuance.
Amendment No.: 167.
Facility Operating License No. NPF-49: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 30, 1998 (63
FR 71971).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
ThreeRivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
PP&L, Inc., Docket No. 50-388, Susquehanna Steam Electric Station, Unit
2, Luzerne County, Pennsylvania
Date of application for amendment: August 5, 1998, as supplemented
by letter dated November 23, 1998.
Brief description of amendment: This amendment would change the
allowable values for both the core spray system and the low-pressure-
coolant injection system reactor steam dome pressure-low functions.
Date of issuance: March 4, 1999.
Effective date: As of date of issuance, to be implemented within 30
days.
Amendment No.: 155.
Facility Operating License No. NPF-22: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 1, 1999 (64 FR
4904).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 4, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek
Generating Station, Salem County, New Jersey
Date of application for amendment: August 25, 1998, as supplemented
January 27, 1999.
Brief description of amendment: This amendment revised Technical
Specification (TS) 2.1.2, ``THERMAL POWER, High Pressure and High
Flow,'' and the Bases for TS 2.1, ``Safety Limits.'' These changes were
made to implement appropriately conservative Safety Limit Minimum
Critical Power Ratio values for the Hope Creek Generating Station Cycle
9 core and fuel designs. An administrative revision has also been made
to TS 6.9.1.9 to reflect these changes for Cycle 9.
Date of issuance: March 9, 1999.
Effective date: As of the date of issuance, to be implemented
within 60 days after the completion of Cycle 8.
Amendment No.: 117.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 23, 1998 (63
FR 50938).
The supplemental letter provided clarifying information that did
not change the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 9, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, NJ 08070.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of application for amendments: September 29, 1998.
Brief description of amendments: The amendments revise Technical
Specification (TS) 3/4.9.4, ``Refueling Operations--Containment
Building Penetrations,'' to allow the use of an equivalent closure
device to satisfy the closure requirements of the containment equipment
hatch during core alterations or movement of irradiated fuel in
containment. The amendment also revises TS 3/4.9.4 to allow the use of
an equivalent closure method to satisfy the closure requirements of
containment penetrations (in addition to an isolation valve, blind
flange or manual valve) during core alterations or movement of
irradiated fuel in containment.
Date of issuance: February 26, 1999.
Effective date: Effective as of its date of issuance, to be
implemented within 60 days.
Amendment Nos.: 217 and 199.
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 21, 1998 (63 FR
56258).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 26, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna
Nuclear Power Plant, Wayne County, New York
Date of application for amendment: November 24, 1998.
Brief description of amendment: This amendment revises the Ginna
Station Improved Technical Specifications
[[Page 14291]]
description of the fuel cladding material (TS 4.2.1) and updates the
list of references provided in Specification 5.6.5 for the Core
Operating Limits Report.
Date of issuance: March 3, 1999.
Effective date: As of date of issuance, to be implemented within 30
days.
Amendment No.: 73.
Facility Operating License No. DPR-18: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 30, 1998 (63
FR 71972).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610.
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 7, 1998, as supplemented by letters
dated October 15 and October 26, 1998, and February 16, 1999. The
supplements provided clarifying information and corrected
administrative errors within the scope of the amendment request and did
not change the initial no significant hazards consideration
determination.
Brief description of amendments: The amendments revised the spent
fuel pool criticality analysis and rack utilization schemes by allowing
credit for spent fuel pool soluble boron.
Date of issuance: March 3, 1999.
Effective date: This license amendment is effective as of its date
of issuance and shall be implemented within 90 days of issuance.
Amendment Nos.: Unit 1--Amendment No. 104; Unit 2--Amendment No.
91.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 26, 1998 (63 FR
45530).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 3, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear
Plant, Unit 2, Limestone County, Alabama
Date of application for amendment: September 8, 1998 (TS-354), as
supplemented by letter dated February 22, 1999.
Brief description of amendment: Revises the Appendix A Technical
Specifications (TS) to include provisions for enabling the Oscillation
Power Range Monitor Upscale trip function in the Average Power Range
Monitor.
Date of issuance: As of date of issuance to be implemented at the
end of the Unit 2 Cycle 10 outage scheduled to begin on April 11, 1999.
Effective date: March 5, 1999.
Amendment No.: 258.
Facility Operating License No. DPR-52: Amendment revises the TS.
Date of initial notice in Federal Register: October 7, 1998 (63 FR
53958). The supplemented letter dated February 22, 1999, did not change
the original no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 5, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of application for amendments: November 10, 1998.
Brief description of amendments: The amendments revise the
Technical Specifications Sections 3.4.4 and 3.4.4.a for Unit 1, and
3.4.4 and 3.4.4.a for Unit 2, providing a clarification on the
operability requirements for pressurizer heaters and the emergency
power source for the pressurizer heaters.
Date of issuance: March 1, 1999.
Effective date: March 1, 1999.
Amendment Nos.: 217 and 198.
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: December 2, 1998 (63 FR
66605).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of application for amendments: October 25, 1995, as
supplemented February 5, 1999. The February 5, 1999, supplemental
letter contained clarifying information only, and did not change the
initial no significant hazards consideration determination or expand
the scope of the original Federal Register Notice.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) Sections 3.4.3.2, 4.4.3.2.1.b,
4.4.3.2.1.c, 4.4.3.2.2, 4.4.9.3.d, 4.4.9.3.e, 3/4.4.2, 3/4.4.3, 3/4.4.4
and 6.8.4.g for Unit 1, and 3.4.3.2, 4.4.3.2.1.c, 4.4.3.2.2, 4.4.9.3.d,
4.4.9.3.e, 3/4.4.2, 3/4.4.3, 3/4.4.4 and 6.8.4.g for Unit 2, providing
an allowed outage time of 14 days for the prezzurizer power operated
relief valve (PORV) nitrogen accumulators, as well as provide separate
action statements for the PORV depending on the reason for the PORV
inoperability.
Date of issuance: March 2, 1999.
Effective date: March 2, 1999.
Amendment Nos.: 218 and 199.
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
28620).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 2, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: September 24, 1998.
Brief description of amendments: These amendments revise the
Technical Specifications to allow the reactor trip bypass breakers to
be tested immediately after being placed in service, but prior to
commencing Reactor Protection System testing or maintenance.
Date of issuance: March 12, 1999.
Effective date: March 12, 1999.
Amendment Nos.: 219 and 219.
Facility Operating License Nos. DPR-32 and DPR-37: Amendments
change the Technical Specifications.
[[Page 14292]]
Date of initial notice in Federal Register: February 10, 1999 (64
FR 6715).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 12, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: September 28, 1998 (TSCR 208).
Brief description of amendments: These amendments clarify the
notation definition of refueling interval ``R'' in TS Table 15.4.1-1
and add a new annual (12-month) interval ``A''.
Date of issuance: March 1, 1999.
Effective date: March 1, 1999, with full implementation within 45
days.
Amendment Nos.: 186 and 191.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 27, 1999 (64 FR
4162).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 1, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: October 5, 1998 (TSCR 200).
Brief description of amendments: These amendments modify TS Section
15.4.1, ``Operational Safety Review,'' by removing the requirement to
check certain environmental monitors on a monthly basis.
Date of issuance: March 2, 1999.
Effective date: March 2, 1999, with full implementation within 45
days.
Amendment Nos.: 187 and 192.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 27, 1999 (64 FR
4163).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 2, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: October 7, 1998 (TSCR 207).
Brief description of amendments: These amendments incorporate
changes to the Technical Specifications to ensure the 4 kV bus
undervoltage input to the reactor trip protective function is
controlled in accordance with the design and licensing basis for the
facility. An additional administrative change removes the footnote
related to the definition of Rated Power in TS 15.1.j.
Date of issuance: March 2, 1999.
Effective date: March 2, 1999, with full implementation within 45
days.
Amendment Nos.: 188 and 193.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 30, 1998 (63
FR 71978).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 2, 1999.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Lester Public Library,
1001 Adams Street, Two Rivers, Wisconsin 54241.
Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power
Station, Franklin County, Massachusetts
Date of application for amendment: October 15, 1998.
Brief description of amendment: Revises the Possession Only License
by changing the submittal interval for the Radioactive Effluent Reports
from semiannual to annual.
Date of issuance: March 5, 1999.
Effective date: March 5, 1999.
Amendment No.: 151.
Possession Only License No. DPR-3: Amendment revised the Technical
Specifications.
Date of initial notice in Federal Register: November 18, 1998 (63
FR 64128). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated No significant hazards
consideration comments received: No.
Local Public Document Room location: Greenfield Community College,
1 College Drive, Greenfield, Massachusetts 01301.
Dated at Rockville, Maryland, this 17th day of March 1999.
For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 99-7032 Filed 3-23-99; 8:45 am]
BILLING CODE 7590-01-P