[Federal Register Volume 59, Number 206 (Wednesday, October 26, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-11026]
[[Page Unknown]]
[Federal Register: October 26, 1994]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating
LicensesInvolving 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 September 30, through October 14, 1994. The
last biweekly notice was published on October 12, 1994 (59 FR 51616).
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 Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, 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
20555. The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By November 25, 1994, 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 20555 and at the local
public document room for the particular facility involved. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 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 20555, and at the local public document
room for the particular facility involved.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: September 23, 1994
Description of amendments request: The proposed amendments would
revise the Unit 2 Shutdown AC Sources Technical Specifications (TSs) to
allow a one-time extension from 7 to 14 days of the allowed outage time
(AOT) for the dedicated Class 1E emergency power source during the
upcoming Unit 2 1995 Refueling Outage (RFO-10). The proposed amendments
would also revise the Unit 1 Control Room Emergency Ventilation System
(CREVS) TSs to provide a one-time extension from 7 to 30 days of the
AOT for one train of the CREVS to be inoperable. As noted, these
extensions will be needed during the upcoming 1995 Unit 2 RFO-10 to
support the modifications scheduled for the onsite electrical
distribution system in response to the Station Blackout (SBO) Rule, 10
CFR 50.63, and the upgrade of No. 21 Emergency Diesel Generator
(EDG).The specific changes requested are:
Unit 2 TSs 3.8.1.2 and 3.8.2.2 will include a footnote indicating
that the AOT for aligning an operable emergency diesel generator (EDG)
to provide power to the emergency busses within 14 days during the Unit
2 RFO-10.
Unit 1 TS 3.7.6.1 will be modified to indicate that during the No.
21 EDG upgrade, the time to restore the No. 21 filter train of the air
conditioning unit to operable status may be extended to 30 days (for
loss of emergency power only) if: 1) A temporary diesel generator is
demonstrated to be available by starting it at least once per 7 days
and 2) if action 1 is not met, restore compliance with the action
within 7 days or be at least in hot standby within the next 6 hours and
in cold shutdown within the following 30 hours.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of issue of no significant hazards consideration
for each of the proposed changes, which is presented below:
In relation to the requested changes to the Unit 2 TSs:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
Requiring one Class 1E Emergency Diesel Generator (EDG) to be
available for a shutdown unit ensures that AC power will be
available for a loss of offsite power event, a boron dilution event,
or a fuel handling incident. There is a very low probability that a
loss of offsite power will occur due to severe weather or
inadvertent damage to the switchyard during the 14-day period that
the temporary splice box is being installed and No. 12 EDG is out-
of-service. The Calvert Cliffs offsite power supply is highly
redundant and has significant capability in withstanding severe
weather events, such as tornadoes. In addition, Calvert Cliffs
Emergency Response Plan Implementation Procedures requires that
certain actions be taken, up to and including shutdown of both
units, on the approach of a severe storm, such as a hurricane. The
probability of a loss of offsite power is maintained low by
prohibiting planned maintenance on two of the three 500 kV
transmission lines and associated relaying and devices within the
switchyard. Availability of the required offsite power sources will
be verified once per shift. In addition to the offsite power
sources, a temporary diesel generator will also be installed to
provide a backup onsite power source with the capacity to support
the safety-related loads of the shutdown unit.
The boron dilution event and the fuel handling incident are the
only two accidents that are explicitly analyzed in the Updated Final
Safety Analysis Report for a shutdown unit. The potential accident
precursors such as core alterations, positive reactivity insertions,
movement of irradiated fuel and movement of heavy loads over
irradiated fuel, will be prohibited while No. 12 EDG is out-of-
service for the temporary splice box installation. Therefore the
probability of a boron dilution event or fuel handling incident is
decreased during the operations allowed by this change. The
requirement to maintain containment penetration closure ensures that
the consequences of an accident would not be significantly
increased.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
A temporary diesel generator is being installed onto a 4 kV bus
of the shutdown unit while the dedicated EDG for this unit is
transferred to the operating unit for up to 14 days. This is an
extension of the same configuration allowed by Action Statements
3.8.1.2.b and 3.8.2.2.b with additional provision taken for the
Control Room Emergency Ventilation System (CREVS). The EDGs will be
aligned so that each train of the CREVS will have an emergency power
supply available. The proposed change has been evaluated and it has
been determined that it does not impair any existing safety-related
equipment needed to maintain the unit in a safe shutdown condition,
and does not create any new accident initiators. The operation of
the temporary diesel generator is familiar to the operators and is
not significantly different from typical operator activities.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The safety function provided by the AC electrical power sources
and associated distribution systems for a shutdown unit is to ensure
that the unit can be maintained in a safe shutdown condition, and
there is sufficient instrumentation and control capability available
for monitoring and maintaining the unit status. The proposed change
would allow the shutdown unit to be without a dedicated Class 1E
emergency power source for up to 14 days. This is an extension of
the outage time of seven days allowed by the Technical
Specifications for performing maintenance and inspections on No. 12
EDG. This proposed change will have no impact on the offsite power
sources.
Several compensatory measures will be taken during this period
to ensure that a power source will be available for the shutdown
unit. These measures include requiring that two offsite power
sources are available, and a temporary diesel generator will be
installed capable of supplying the loads necessary to maintain the
unit in a safe condition. In addition, Technical Specifications
require several compensatory measures to reduce the potential for a
fuel handling incident and a boron dilution event. These measures
include prohibiting positive reactivity changes, suspending core
alterations, movement of irradiated fuel, and the movement of heavy
loads over irradiated fuel. Establishing containment penetration
closure further ensures that adequate margin of safety is
maintained. In addition, reduced inventory conditions of the Reactor
Coolant System will be prohibited during the 14-day period.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
In relation to the requested changes in the Unit 1 TSs:
1. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The Control Room Emergency Ventilation System (CREVS) is
designed so that the Control Room can be occupied under all plant
conditions. The CREVS is required to maintain the Control Room
temperature and to filter the Control Room air in the event of a
radioactive release. When No. 21 Emergency Diesel Generator (EDG) is
being upgraded, No. 12 CREVS will be without a Class 1E emergency
power source. The CREVS is not an initiator in any previously
evaluated accidents. Therefore, the proposed change does not involve
an increase in the probability of an accident previously evaluated.
The CREVS is required to maintain the Control Room habitable
following a radioactive release from a loss of coolant accident, a
main steam break, or a steam generator tube rupture. There is a very
low probability of an event occurring requiring Control Room
isolation during the 30-day period that it will take to upgrade No.
21 EDG. Requiring that the CREVS have both a normal power source and
an emergency power source available ensures that one train of the
system will be available so that the Control Room can be occupied
under these conditions. The probability of a loss of offsite power
is very low due to the highly redundant design of the offsite power
supply. Planned maintenance on three of the offsite power supplies
and associated relaying and devices within the switchyard will be
prohibited during the upgrade period to maintain the low probability
of a loss of offsite power event. Number 12 CREVS train will
continue to have its normal power source for all but approximately
four days when the bus will be de-energized to allow bus work that
is necessary to the tie-in of the Alternate AC diesel generator.
Number 11 CREVS will have both its normal and emergency power supply
available and this train is capable of maintaining the Control Room
habitable. In addition, a temporary diesel generator will be
installed to provide assurance that an emergency power source will
be available to No. 12 CREVS. The compensatory measures that will be
taken during this period will ensure that the proposed change does
not involve a significant increase in the consequences of an
accident previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The CREVS is not being modified by this proposed change. The
system will continue to operate in the same manner. Number 21 EDG
will operate in a similar manner after the upgrade and will be able
to support unit operation after all the testing is completed. The
installation of the temporary diesel generator during the upgrade
period has been evaluated to ensure that it does not create any new
accident initiators.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
The operability of the CREVS during Modes 1 through 4 ensures
that the Control Room will remain habitable under all plant
conditions. The proposed change does not affect the function of the
CREVS. The proposed change will allow one train of the CREVS to be
without a Class 1E emergency power supply for up to 30 days. This
train will have the normal power supply available for all but
approximately four days to allow necessary bus work. The other train
of the CREVS will have both its normal and emergency power supplies
during this period. Compensatory measures that will be taken include
prohibiting planned maintenance on the required offsite power
sources and installing a temporary diesel generator of sufficient
capacity as a backup to the affected train. These measures will
maintain the current margin of safety. The upgrade to the existing
EDGs will provide additional margin for the electrical loading of 4
kV safety-related busses. The completion of the No. 21 EDG upgrade
will improve the margin of safety for the onsite electrical
distribution system.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Ledyard B. Marsh
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: September 6, 1994
Description of amendment request: The proposed amendment changes
the Pilgrim Nuclear Power Station Technical Specifications Sections
3.7.B.1.a, 3.7.B.1.c, 3.7.B.1.e, 3.7.B.2.a, and 3.7.B.2.c. The proposed
changes also add new sections 3.7.B.1.f and 3.7.B.2.e. These sections
require both trains of the Standby Gas Treatment (SGTS) and Control
Room High Efficiency Air Filtration (CRHEAF) System to be operable for
the initiation of fuel movement and during fuel handling operations
involving irradiated fuel.
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 operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Technical Specifications 3.7.B.1 and 3.7.B.2.e restrict the
movement of irradiated fuel when only one train of SGTS or one train
of CRHEAF are operable. Irradiated fuel movement may not begin and
may only continue for seven days when the Limiting Condition of
Operation is entered.
Removing these restrictions during refueling operations does not
involve a significant increase in the probability or consequences of
an accident previously evaluated because compensatory measures will
be in place.
When sections 3.7.B.1.f and 3.7.B.2.e are invoked fuel movement
will not commence until 5 days following plant shutdown and reactor
vessel will be flooded-up to elevation 114''. The 5 day period
provides decay-time before irradiated fuel movement begins.
Flooding-up elevation 114'' provides an enlarged inventory reducing
the possibility of a loss-of-coolant event exposing fuel such that
radioactive gasses are produced, an event SGTS and CRHEAF are
designed to mitigate.
Other compensatory measures include requiring the SBO [station
blackout] diesel or the shutdown transformer to be operable prior to
and during the fuel movement. This adds defense-in-depth by making
available another power supply to the in-service safety-related bus.
Also, the substitution of a non-safety power supply to the SGTS and
CRHEAF ``inoperable'' systems while their safety-grade bus is out-
of-service for maintenance will provide offsite power to the
``inoperable'' train. While this electrical supply is not safety-
grade, it is reliable and capable of powering the SGTS and CRHEAF
systems. The components of the ``inoperable'' trains will be
available with power from an alternate power source. The
compensatory connection to the non-safety grade bus gives added
confidence these trains can perform the design function although
they are not ``operable'' as defined by Technical Specifications.
Operating Pilgrim in accordance with this proposed change does
not involve a significant increase in the probability or consequence
of an accident previously analyzed because compensatory measures
will be in force to: restrict the commencement of irradiated fuel
handling or new fuel handling over the spent fuel or core until 5
days following reactor shutdown; provide a reliable source of power
to the ``inoperable'' SGTS and CRHEAF systems; provide an enlarged
coolant inventory to protect irradiated fuel from the effects of an
inadvertent draindown of the vessel; and provide an additional
source of emergency power to the active SGTS and CRHEAF systems by
ensuring the operability of the SBO diesel generator or the Shutdown
Transformer.
2. The operation of Pilgrim Station in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Planned maintenance activities require removing a safety-related
bus and emergency diesel generator powering a train of SGTS and
CRHEAF from service. The redundant trains are not affected. The
affected trains of SGTS and CRHEAF will be connected to a non-safety
bus, allowing them to operate but not allowing them to be considered
operable under the purview of Technical Specifications. The proposed
change allows refueling activities to commence with one train of
SGTS and CRHEAF fully operable and the other train available but not
powered by its safety grade bus and associated emergency diesel
generator. Compensatory measures will be in effect during refueling
activities involving this configuration. The proposed changes do not
create the possibility of a new or different kind of accident from
the fuel-drop accident previously analyzed. Therefore, operating
Pilgrim in accordance with this change will not create the
possibility of a new or different kind of accident from any accident
previously analyzed.
3. The operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant reduction in the
margin of safety.
SGTS and CRHEAF contribute to the margin of safety during fuel
handling by mitigating the consequences of a fuel-handling event.
Allowing an exception to the requirement of both trains of SGTS and
CRHEAF operable prior to or during fuel movement activities does not
involve a significant reduction in the margin of safety because the
first line of defense, the other SGTS and CRHEAF trains, will be
operable. The redundant trains will also be powered and operable in
all ways except the ``operable'' concept required by Technical
Specification.
Hence, the actual condition of the equipment allows it to meet
its design function except under the strict Technical Specification
interpretation of operable, and the described compensatory measures
that will be in effect when the exception is employed, constrain the
potential impact on the margin of safety caused by using the
exception; therefore, operating Pilgrim in accordance with this
proposed Technical Specification request 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 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: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
NRC Project Director: Walter R. Butler
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: September 6, 1994
Description of amendment request: The proposed amendment would
reduce the Reactor Pressure Setpoint at which the shutdown cooling
system automatically isolates. This setpoint also isolates the low
pressure coolant injection valves when the shutdown cooling system is
in operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Technical Specification Table 3.2.A lists the instrumentation
that initiates primary containment isolation and also lists the trip
level setting (setpoints) for that instrumentation. The setpoint for
reactor high pressure is presently [less than or equal to] 110 psig
which was selected to provide protection for the RHR [residual heat
removal] low pressure suction piping against possible
overpressurization. This signal initiates a group 3 containment
isolation by closing the shutdown cooling isolation valves and the
Low Pressure Coolant Injection (LPCI) valves. To provide an optimal
solution to address Generic Letter 89-10, the motor-operated valves
which effect the isolation of the RHR suction piping (MO1001-47 and
MO1001-50) are being modified based on a lower differential pressure
in the design calculations. The setpoint is being reduced to ensure
plant operation is maintained in accordance with the new design and
to continue to provide the protection necessary against
overpressurization. This does not involve an increase in the
probability or consequences of an accident previously analyzed
because reducing the setpoint to less than what the technical
specifications currently requires is a change in the conservative
direction relative to protection of the piping. The LPCI injection
valves are designed for higher pressures and the proposed setpoint
change does not involve an increase in the probability or
consequences of an accident previously evaluated.
Technical Specification Table 3.2.B lists instrumentation that
initiates or controls the core and containment cooling systems and
also lists the trip level settings (setpoints) for that
instrumentation. The setpoint for reactor low pressure [less than or
equal to] 110 psig, is a permissive for the group 3 isolation of the
RHR inboard injection valves. Reducing the setpoint to [less than or
equal to] 76 psig is consistent with the design of the other group 3
isolation valves that receive the same signal and accomplishes the
isolation of the shutdown cooling system when there is a system
breach. Thus, revising this setpoint does not increase the
probability or consequences of an accident previously evaluated.
2. The operation of Pilgrim Station in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed setpoint change supports modifications made to the
shutdown cooling isolation valves to provide additional margin to
address Generic Letter 89-10 concerns. Reducing the setpoint for
this function continues to provide protection of the RHR suction
piping and ensures closure of the isolation valves. Therefore,
revising the reactor high pressure setpoint to [less than or equal
to] 76 psig for instrumentation that initiates primary containment
isolation (Table 3.2.A) does not create the possibility of a new or
different kind of accident previously evaluated. Similarly, the
revision of the reactor low pressure setpoint to [less than or equal
to] 76 psig for instrumentation that initiates or controls the core
and containment cooling systems does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant reduction in a
margin of safety.
The purpose of the setpoint for reactor pressure in Table 3.2.A
and 3.2.B is to provide protection for the RHR suction piping and
ensure proper isolation for unlikely piping breaches. Changing the
setpoint to a lower value is consistent with modifications being
made to the shutdown cooling isolation valves. The margin of safety
for this setpoint was established to protect the RHR suction piping
from overpressurization and to ensure that primary containment
integrity could be established by the isolation valves on a Group 3
isolation. A margin of safety for protecting the RHR suction piping
exists due to the difference between the design pressure of the
piping and the setpoint specified in the technical specifications.
Reducing the setpoint increases the difference between the design
pressure of the piping and the setpoint hence, this margin of safety
is increased. The margin of safety established for primary
containment isolation valves is maintained by specifying a setpoint
which corresponds to the closing differential pressure of the valves
under postulated accident conditions. The setpoint change does not
reduce the design margins established to ensure the valves perform
their design isolation function when required. The low pressure
coolant injection valves that receive this same signal are designed
for higher pressures than the current setpoint of [less than or
equal to] 110 psig and, therefore, a lower setpoint increases the
margin of safety. Thus, the proposed amendment does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
NRC Project Director: Walter R. Butler
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: September 6, 1994
Description of amendment request: The proposed amendment would
remove Technical Specification section 4.5.H.4, a section which
requires the testing and calibration of pressure switches in certain
emergency core cooling system (ECCS) lines.
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 Operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
[***].
The discharge piping for ECC systems is maintained filled to
prevent water hammer during automatic pump starts. Monthly venting
is the primary means of ensuring filled discharge piping. The
pressure switches are an adjunct to such venting. Hence, piping in
the Core Spray System, the Low Pressure Coolant Injection System
(LPCI), the High Pressure Coolant Injection (HPCI) system, and the
Reactor Core Isolation Coolant (RCIC) system are all equipped with
pressure switches that detect pressure decay in the discharge piping
of these systems.
This proposed change does not change Pilgrim's configuration or
equipment. The switches perform a surveillance function and do not
provide a signal needed to prevent or mitigate an accident. The
switches will continue to perform their surveillance function and
their surveillance and calibration will be performed in accordance
with Pilgrim procedures. Removal of section 4.5.H.4 eliminates the
possibility of inoperable switches forcing the shutdown of Pilgrim
or the alternative of declaring an operable safety system inoperable
because of its association with these switches.
Technical Specifications will continue to require venting the
discharge piping high point when the systems are configured such
that water hammer can occur. (sections 4.5.H.1, 4.5.H.2 and
4.5.H.3). Thus, the application of this proposed change does not
reduce the Technical Specifications intent of reducing the
likelihood of discharge piping water hammer. Therefore, operating
Pilgrim Station in accordance with the proposed amendment will not
involve a significant increase in the probability or consequences of
an accident previously analyzed.
2. The operation of Pilgrim Station in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Section 4.5.H's purpose is to maintain the ECCS discharge piping
filled to prevent water hammer. The purpose of the pressure switches
is to detect voids in ECCS discharge piping to prevent the
possibility of damage due to water hammer. These switches are not
safety-related, have no automatic functions, and are not relied on
to prevent or mitigate an accident. Instead, they enhance the
existing discharge pipe venting surveillance requirements by
detecting void formation in discharge pipe.
The switches will continue to perform their surveillance
function through Pilgrim procedures. Venting will continue to be
required by Technical Specifications. Therefore, operating Pilgrim
in accordance with this proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated because the proposed change does not impair the
detection of conditions necessary to produce a water hammer in the
discharge piping.
3. The operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant reduction in a
margin of safety.
The discharge piping pressure switches are surveillance
instruments and act as a secondary means of protecting the discharge
piping from conditions that can produce water hammer. They are not
relied on to prevent or mitigate accidents. Hence, these switches do
not significantly impact safety because they are not the primary
means of preventing discharge piping water hammer. Therefore,
removing the pressure switches from Technical Specifications
potentially contributes to plant availability but does not involve a
significant reduction in a margin of safety because the primary
method of detection (venting) remains and the switches will continue
to be subject to procedural controls.
This proposed change has been reviewed and recommended for
approval by the Operations Review Committee and reviewed by the
Nuclear Safety Review and Audit Committee.
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: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
NRC Project Director: Walter R. Butler
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: September 6, 1994
Description of amendment request: The proposed amendment would
relocate the alarms for the drywell to suppression chamber vacuum
breakers to a different annunciator panel.
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 operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously identified.
The proposed change relocates annunciators in the control room
but does not change their designed function or setpoint.
The Annunciator System is non-safety related and performs no
direct safety function. No accident initiators are being affected by
this proposed change. Accident mitigating systems remain operable,
and accident scenarios are unaffected.
2. The operation of Pilgrim Station in accordance with the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously analyzed.
Relocating the Drywell to Suppression Chamber annunciator from
one control room panel to another does not create the possibility of
a new or different kind of accident. This modification does not
modify the setpoints or functions of the annunciators. Hence, it is
administrative and proposed to allow relocation which is currently
constrained by the current Technical Specifications level of detail.
3. The operation of Pilgrim Station in accordance with the
proposed amendment will not involve a significant reduction in the
margin of safety.
The equipment being relocated is non-safety related and its
relocation does not impact the margin of safety. This relocation is
proposed to enhance the operator's ability to identify and analyze
abnormal events.
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: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
NRC Project Director: Walter R. Butler
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois;Docket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of amendment request: June 13, 1994, as supplemented on
October 7, 1994.
Description of amendment request: The proposed amendment would make
several changes to the Administrative Controls in Section 6 of
Technical Specifications (TS) for Byron and Braidwood stations. The
proposed changes include: (1) a change to the submittal frequency of
the Radiological Effluent Release Report, (2) a revision to the Shift
Technical Advisor description, (3) clarification of the Shift
Engineer's responsibilities, and (4) editorial changes. The references
to the Semiannual Radiological Effluent Release Report are also revised
in other sections of the TS. The proposed change in the October 7,
1994, submittal revised TS 6.3.1 to include generic descriptions of
personnel who fulfill the responsibilities of a radiation protection
manager. This supplements the information that was published in the
Federal Register on August 3, 1994 (59 FR 39581).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
A. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes to Section 6 of Technical Specifications do
not affect any accident initiators or precursors and do not change
or alter the design assumptions for the systems or components used
to mitigate the consequences of an accident.
The proposed changes are administrative in nature and provide
clarification. These changes provide consistency with station
procedures, programs, the Code of Federal Regulations, other
Technical Specifications, and Standard Technical Specifications.
These changes do not impact any accident previously evaluated in the
Updated Final Safety Analysis Report.
B. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not affect the design or operation of
any system, structure, or component in the plant. There are no
changes to parameters governing plant operation; no new or different
type of equipment will be installed. The proposed changes are
considered to be administrative changes. All responsibilities
described in Technical Specifications for management activities will
continue to be performed by qualified individuals.
C. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes do not affect the margin of safety for any
Technical Specification. The initial conditions and methodologies
used in the accident analyses remain unchanged, therefore, accident
analysis results are not impacted.
The proposed changes are administrative in nature and have no
impact on the margin of safety of any Technical Specification. They
do not affect any plant safety parameters or setpoints. The
descriptions for the Shift Technical Advisor and Shift Engineer are
clarified, however, include no reduction to their responsibilities.
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: For Byron, the Byron Public
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee
Street, Wilmington, Illinois 60481.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of amendment request: September 13, 1994
Description of amendment request: The amendments replace
Containment Systems technical specification (TS) 3.6.2.2, ``Spray
Additive System'' with a new Emergency Core Cooling Systems TS 3.5.5,
``ECCS Recirculation Fluid pH Control System.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed change involves replacement of concentrated
NaOH injected via the containment spray system with trisodium
phosphate (TSP) stored in the containment and dissolved in the sump
recirculation solution to maintain acceptable post accident spray/
recirculation solution chemistry. Deletion of the concentrated NaOH
will eliminate a personnel hazard. The pH control system functions
in response to an accident and does not involve or have any effect
on any initiating event for any accident previously evaluated.
Operation under the proposed amendment will continue to ensure that
iodine potentially released post-LOCA is retained in the sump
solution, and resultant offsite and control room thyroid doses are
within the limits of 10 CFR 100 and 10 CFR 50, Appendix A, General
Design Criterion 19, respectively.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The deleted equipment is isolated from the remaining
equipment by blind flanges, locked closed valves, cut and capped
piping, determinated and/or spared cables; and interfaces are
analyzed to ensure the remaining required equipment meets applicable
original design requirements. The new equipment (TSP and baskets) is
a passive pH control system and is supported and analyzed to ensure
there are no adverse interfaces (e.g. pipe break, jet impingement,
seismic) with existing equipment, systems, or structures.
3. The proposed change does not involve a significant reduction
in a margin of safety. The slight change in recirculation solution
pH maintains adequate protection against chloride induced stress
corrosion cracking of austenitic stainless steel and maintains the
capability of the solution to retain iodine. It results in an
insignificant increase in the post-accident rate of hydrogen
generation, which remains well within the existing capacity of the
hydrogen recombiners. The increased mass in the containment will
have no significant impact on post-accident flood levels,
recirculation solution boron concentration, or peak clad
temperatures. No other operating parameters for systems, structures,
or components assumed to operate in the safety analysis are changed.
The offsite and control room doses meet the limits of 10 CFR 100 and
GDC 19 respectively. Because the trisodium phosphate is nonvolatile
and the baskets are protected with solid covers and are located
slightly above the floor in the containment where access is strictly
controlled, a surveillance interval of once per refueling outage
provides assurance that the TSP will be available when required.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia 30830.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308
NRC Project Director: Herbert N. Berkow
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: September 26, 1994
Description of amendment request: The proposed license amendment
would revise the ``Plan for the Long Range Planning Program'' by
changing the semi-annual reporting period to annual, and to reflect
refined evaluation criteria and assessment methodology; and, to
incorporate the necessary changes to the license condition wording.
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 of occurrence or consequences of an accident previously
evaluated. The proposed revision to the Facility Operating License
does not affect the safety analysis and does not involve any
physical changes to the plant, nor any changes in the format or
restraints on plant operations, and only contemplates a change to
the Plan for the Long Range Planning Program currently approved by
the NRC in license condition 2.C.(6). Therefore, this change will
not increase the probability of previously analyzed accidents
because it involves no direct plant modification or change in
operation, and hence, it is also unrelated to the possibility of
increasing the consequences of previously analyzed accidents.
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. The proposed
revision to the Facility Operating License does not affect the
safety analysis and does not involve any physical changes to the
plant, nor any changes in the format or restraints on plant
operations, and only contemplates a change to the Plan for the Long
Range Planning Program currently approved by the NRC in license
condition 2.C.(6). Therefore, this change has no effect on the
possibility of creating a new or different kind of accident from any
previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. The proposed revision to the Facility Operating License does
not involve any physical changes to the plant, nor any changes in
the format or restraints on plant operations, and only contemplates
a change to the Plan for the Long Range Planning Program currently
approved by the NRC, in license condition 2.C.(6). Therefore, the
overall margin of safety for the plant is maintained.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Phillip F. McKee
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: September 26, 1994
Description of amendment request: The proposed license amendment
would revise the ``Plan for the Long Range Planning Program'' by
changing the semi-annual reporting period to annual, and to reflect
refined evaluation criteria and assessment methodology; and, to
incorporate the necessary changes to the license condition wording.
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 of occurrence or consequences of an accident previously
evaluated. The proposed revision to the Facility Operating License
does not affect the safety analysis and does not involve any
physical changes to the plant, nor any changes in the format or
restraints on plant operations, and only contemplates a change to
the Plan for the Long Range Planning Program currently approved by
the NRC in license condition 2.C.(9). Therefore, this change will
not increase the probability of previously analyzed accidents
because it involves no direct plant modification or change in
operation, and hence, it is also unrelated to the possibility of
increasing the consequences of previously analyzed accidents.
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. The proposed
revision to the Facility Operating License does not affect the
safety analysis and does not involve any physical changes to the
plant, nor any changes in the format or restraints on plant
operations, and only contemplates a change to the Plan for the Long
Range Planning Program currently approved by the NRC in license
condition 2.C.(9). Therefore, this change has no effect on the
possibility of creating a new or different kind of accident from any
previously evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety. The proposed revision to the Facility Operating License does
not involve any physical changes to the plant, nor any changes in
the format or restraints on plant operations, and only contemplates
a change to the Plan for the Long Range Planning Program currently
approved by the NRC, in license condition 2.C.(9). Therefore, the
overall margin of safety for the plant is maintained.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, 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: Phillip F. McKee
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: September 26, 1994
Description of amendment request: The proposed amendment would
revise the Cooper Nuclear Station (CNS) Technical Specifications,
Section 3.5.C ``HPCI System,'' to increase the minimum pressure at
which the High Pressure Coolant Injection (HPCI) System is required to
be OPERABLE from greater than 113 psig to greater than 150 psig.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Evaluation
The change in the reactor vessel pressure at which the High
Pressure Coolant Injection (HPCI) System must be operable from
113 psig to 150 psig will not result in a significant increase
in the probability or consequences of an accident previously
evaluated. The HPCI System is designed to provide adequate reactor
vessel coolant injection for small break accidents where the reactor
vessel remains pressurized. Therefore, the HPCI System provides a
means of responding to previously analyzed accidents. Changing the
lower bound reactor vessel pressure limit at which the HPCI System
must be operable does not affect any of the accident initiation
sequences previously analyzed, and therefore this proposed change
will not result in an increase in the probability of any accident
previously analyzed.
The change in the required pressure at which the HPCI System
must be operable from 113 psig to 150 psig will not involve a
significant increase in the consequences of any accident previously
evaluated. Increasing this minimum pressure at which the HPCI System
must be OPERABLE will not affect the availability of other systems
which provide standby core cooling. The CNS Core Standby Cooling
Systems (CSCS), which consist of the HPCI System, the Automatic
Depressurization System (ADS), the Low Pressure Coolant Injection
(LPCI) System, and the Core Spray (CS) System, are designed to cover
the spectrum of loss-of-coolant accidents. For large break events,
the reactor vessel will depressurize below the point where the HPCI
System is OPERABLE, and single failure proof core cooling is
provided by a combination of the LPCI and CS systems. For small
break events wherein the reactor vessel does not rapidly
depressurize, the HPCI System is designed to provide core cooling
with a reactor vessel pressure range of 1120 psig to 150 psig. Upon
failure of the HPCI System to provide adequate core cooling, the ADS
in conjunction with the LPCI and CS systems provide single failure
proof assurance of adequate core cooling. The Low Pressure Systems
(LPCI and CS) are designed and required to provide core cooling at
reactor pressures below 150 psig.
The District performed calculations which have determined that
the low pressure Core Standby Cooling systems are capable of
providing adequate core cooling with a reactor pressure of 150 psig
under the most degraded pump conditions, i.e., pump performance at
minimum Technical Specifications requirements. Additionally, the
District reviewed applicable engineering calculations to ensure that
no calculations were relying on the HPCI System to provide degraded
flow to the reactor vessel during any accident scenario or
transient. Based on the diverse means of providing adequate core
cooling for the spectrum of loss-of-coolant accidents, and the
capability of the low pressure core cooling systems to provide
adequate core cooling at 150 psig and below, changing the required
pressure at which HPCI must be operable from 113 psig to 150 psig
will not change the capability to provide adequate core cooling
following postulated events.
The proposed changes do not alter the conditions or assumptions
in any of the Updated Safety Analysis Report (USAR) accident
analyses. Since the USAR accident analyses remain bounding, the
radiological consequences previously evaluated are not adversely
affected by the proposed changes. Therefore, it can be concluded
that the proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2.Does the proposed License Amendment create the possibility of a
new or different kind of accident from any accident previously
evaluated?
Evaluation
The proposed changes introduce no new failure modes for any
plant system or component important to safety nor has any new
limiting failure been identified as a result of the proposed
changes. Increasing the minimum reactor pressure at which the HPCI
System is required to be OPERABLE will not cause an unplanned
initiation of the HPCI System or any other plant system or
equipment, nor will the change impede the initiation of any required
safety system. The HPCI System relies on the containment suppression
pool, emergency condensate storage tanks, plant D.C. electrical
system, and the reactor low water level and high drywell pressure
instrumentation to adequately operate. The proposed increase in the
minimum reactor pressure at which the HPCI System would be required
OPERABLE will not affect the equipment of these systems, nor will
the change affect the physical configuration of the HPCI System.
There will be no change in the types or increase in the amount of
effluents released offsite. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the proposed change create a significant reduction in the
margin of safety?
Evaluation
Changing the reactor vessel pressure at which the HPCI System
must be OPERABLE from 113 psig to 150 psig will not constitute a
significant reduction in the margin of safety. As stated in the
Technical Specifications Bases Section 3.5.C, the HPCI System is
designed to provide rated cooling water flow for reactor pressures
ranging from 1120 psig to 150 psig. The HPCI is not designed to
provide rated cooling water flow at reactor pressures below 150
psig. At reactor operating pressures below 150 psig, the low
pressure core cooling systems are required to be available, are
capable of fulfilling their functions, and provide the required flow
in the low pressure regions below 150 psig. Additionally, the
combination of the ADS, LPCI and CS systems provide additional means
of providing adequate core cooling at any reactor pressure.
Therefore the proposed change to increase the minimum reactor
pressure at which the HPCI System is required to be operable to
greater than 150 psig will not significantly reduce the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Auburn Public Library, 118
15th Street, Auburn, Nebraska 68305
Attorney for licensee: Mr. G.D. Watson, Nebraska Public Power
District, Post Office Box 499, Columbus, Nebraska 68602-0499
NRC Project Director: William D. Beckner
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: September 9, 1994
Description of amendment request: The proposed revision to the
Technical Specifications would delete the requirement for a special
test of the alternate train when one train is inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed changes in accordance with
10CFR50.92 and concludes that the changes do not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
compromised. The proposed changes do not involve an SHC because the
changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed changes do not affect the operation of the APR
[automatic pressure relief] or FWCI [feedwater coolant injection]
subsystems, nor the SBGT [standby gas treatment] system. The
proposed changes do not modify the required actions described in the
LCOs [limiting conditions for operation] when either one or both
circuits of SBGT or an APR valve are determined to be inoperable.
The proposed changes will increase the availability of the APR
subsystem by eliminating a surveillance requirement that causes the
actuation logic to be taken out of service for testing when one
valve is determined to be inoperable. The proposed changes will not
affect the availability of the remaining circuit of SBGT since
testing does not remove the train from service.
Both the SBGT and APR systems function to mitigate the
consequences of postulated accidents. As such, modification to the
surveillance requirements does not create a significant increase in
the probability of an accident. Eliminating the alternate train
testing requirement will not significantly increase the consequences
of a postulated accident. The added assurance that the APR actuation
logic is operable which is provided by Section 4.5.D.2 is not
sufficient to justify the loss of safety function during testing, or
the increased risk of inadvertent operation of the APR valves or the
FWCI subsystem. While Technical Specification 4.7.B.3.c does not
remove the remaining SBGT circuit from service, reasonable assurance
of operability is provided by Technical Specification 4.7.B.2.d
which requires a monthly demonstration of operability of each train
of the SBGT system.
Therefore, no significant increase in the probability or
consequences of an accident previously analyzed would occur.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed changes delete the requirement to demonstrate the
operability of the remaining APR valves actuation logic, the FWCI
subsystem, and the alternate circuit of SBGT immediately and daily
thereafter when one APR valve or one circuit of SBGT is determined
to be inoperable. The proposed changes do not add or change any
equipment or logic. The proposed changes also do not alter any
system operability requirements. These changes only affect the
number of surveillance tests which must be performed. They do not
affect the test methodology for any of these systems.
Since there are no changes to the function, operation, or
surveillance test methodology of any of these systems, the
possibility of a new or different kind of accident is not created.
3. Involve a significant reduction in the margin of safety.
The proposed changes delete the requirement to demonstrate the
operability of the remaining APR valves actuation logic, the FWCI
subsystem, and the alternate circuit of SBGT immediately and
daily thereafter when one APR valve or one circuit of SBGT is
determined to be inoperable. The elimination of the additional
assurance that the actuation logic for the remaining APR valves and
the FWCI subsystem is operable is more than offset by the increase
in the margin of safety which is created by eliminating a
requirement to remove the safety system from service for testing.
The margin of safety for the SBGT system is not significantly
reduced since this system is tested monthly in accordance with
Technical Specification 4.7.B.2.d.
Assurance of operability is provided by the normal, scheduled
surveillances which have been established at a sufficient interval
to provide reasonable assurance of operability. Therefore, the
proposed changes do not involve a significant reduction in the
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
DiabloCanyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of amendment requests: September 20, 1994 (Reference LAR 94-
08)
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant Unit Nos. 1 and 2 to revise surveillance requirements (SRs)
as recommended by NRC Generic Letter (GL) 93-05, ``Line-Item Technical
Specification Improvements to Reduce Surveillance Requirements for
Testing During Power Operation.'' The specific TS changes proposed are
as follows:
(1) TS SR 4.1.3.1.2 would be revised to change the frequency for
testing the movability of the control rods from at least once per 31
days to at least once per 92 days.
(2) TS 3/4.3.2, Table 4.3-2, ``Engineered Safety Features Actuation
System Instrumentation Surveillance Requirements,'' Functional Unit
3.c.4), and TS 3/4.3.3.1, Table 4.3-3, ``Radiation Monitoring
Instrumentation for Plant Operations SRs,'' would be revised to change
the monthly channel functional test to a quarterly channel functional
test.
(3) The proposed changes to TS 3/4.5.1 are as follows: (a)
TS SR 4.5.1.1a.1) would be revised to more clearly state that the
accumulator water volume and pressure must be verified to be within
their limits. (b) TS SR 4.5.1.1b. would be revised to specify that the
boron concentration surveillance is not required to be performed if the
accumulator makeup source was the refueling water storage tank (RWST).
(c) TS SR 4.5.1.2 would be relocated to plant procedures.
(4) TS SR 4.5.2c.2) would be revised to clarify that a separate
containment entry to verify the absence of loose debris is not required
after each containment entry.
(5) TS SR 4.6.2.1d. would be revised to change the frequency for a
containment spray header flow test from at least once per 5 years to at
least once per 10 years.
(6) TS SR 4.6.4.2a. would be revised to change the verification of
the minimum hydrogen recombiner sheath temperature from at least once
per 6 months to at least once each refueling interval.
(7) TS SR 4.7.1.2.1 would be revised to change the surveillance
frequency for testing each auxiliary feedwater (AFW) pump from at least
once per 31 days to at least once per 92 days on a staggered test
basis.
(8) TS SR 4.10.1.2 would be revised to lengthen the allowed period
of time for a rod drop test from 24 hours to 7 days prior to reducing
shutdown margin to less than the limits of TS 3.1.1.1.
(9) TS SR 4.11.2.6 would be revised to change the surveillance
frequency from 24 hours to 7 days when radioactive material is being
added to the gas decay tanks and to add a requirement to monitor
radioactive material concentrations in the gas decay tanks at least
once per 24 hours when system degassing operations are in progress.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The changes proposed in this LAR [license amendment request] are
consistent with the guidance provided in GL 93-05. The proposed
changes eliminate testing that is likely to cause transients or
excessive wear of equipment. An evaluation of these changes
indicates that they result in a net benefit to plant safety. The
evaluation considered:
(i) Unavailability of safety equipment due to testing
(ii) Initiation of significant transients due to testing
(iii) Actuation of engineered safety features that unnecessarily
cycle safety equipment
(iv) Importance to safety of that system or component
(v) Failure rate of that system or component
(vi) Effectiveness of the test in discovering the failure
As a result of the decrease in the testing frequencies, the risk
of testing causing a transient and equipment degradation will be
decreased, and the reliability of the equipment will not be
significantly decreased.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed changes do not affect the method of operating any
equipment at DCPP. Additionally, the proposed changes do not result
in a physical modification to any plant equipment.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
c. Does the change involve a significant reduction in a margin
of safety?
The proposed changes affect the surveillance requirements. There
is no decrease in equipment reliability by the elimination of
unnecessary testing that increases the risk of transients or
equipment degradation.
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 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: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: Theodore R. Quay
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of amendment request: September 19, 1994
Description of amendment request: The proposed license amendment
would revise Technical Specification (TS) 3/4.7.4, ``Snubbers,'' and
its bases, in accordance with NRC Generic Letter (GL) 90-09,
``Alternative Requirements for Snubber Visual Inspection Intervals and
Corrective Actions.'' One difference from GL 90-09 is that the initial
inspection interval using the new criteria would be 18 months from the
conclusion of the visual inspection conducted during the recently
completed refueling outage. Additional changes to the TS would be made
to ensure consistency with the revised snubber visual inspection
interval schedule.
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 change has been reviewed for PNPP and has been
determined not to involve a significant hazards consideration based
on the following:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Implementing the guidance recommended in a GL 90-09 will not
introduce any new failure mode and will not alter any assumptions
previously made in evaluating the consequences of an accident. As
stated in the GL, the proposed alternate schedule for visual
inspections of snubbers will maintain the same operability
confidence level as the existing schedule. Also, the surveillance
requirements and schedule for snubbers functional testing remains
the same, providing a 95 percent confidence level that 90 percent to
100 percent of the snubbers operate within the specified acceptance
limits. The proposed visual inspection schedule is separate from the
functional testing and provides additional confidence that the
installed snubbers will serve their design function and are being
maintained operable. The proposed change does not affect limiting
safety system settings or operating parameters, and does not modify
or add any accident initiating events or parameters. No hardware
modifications are associated with these changes. Therefore, the
proposed change does not significantly increase the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Implementing the recommendations provided in GL 90-09 does not
involve any physical alterations to plant equipment, changes to
setpoints or operating parameters, nor does it involve any accident
initiating event. As stated in the GL, the alternate schedule for
snubber visual inspections maintains the same confidence level as
the existing schedule. In addition to the visual inspections,
functional testing of snubbers, which provides a 95 percent
confidence level that 90 percent to 100 percent of the snubbers
operate within specified acceptance limits, will continue to be
performed. Since this TS change does not physically alter the plant
equipment and the snubber confidence level remains the same there
will not be any new or different accident resulting from snubber
failure from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change incorporates surveillance requirements for
snubber visual inspection intervals that are consistent with the
guidance provided in GL 90-09. As stated in the GL, the proposed
snubber visual inspection interval maintains the same confidence
level as the existing snubber visual inspection interval. This
surveillance requirement does not alter the current Limiting
Condition for Operation or the accompanying actions for the
snubber(s). The requirement for functional testing of safety-related
snubbers is unchanged and remains the basis for the established
margin of safety and assures a 95 percent confidence level that 90
percent to 100 percent of the snubbers operate within the specified
acceptance limits. The functional testing along with the proposed
visual inspection provides adequate assurance that the snubber will
perform its intended function. Therefore, the proposed change does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037NRC Acting Project
Director: Cynthia A. Carpenter
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of amendment request: September 12, 1994
Description of amendment request: The proposed amendment would
modify Point Beach Nuclear Plant Technical Specification (TS) 15.3.3,
``Emergency Core Cooling System, Auxiliary Cooling Systems, Air
Recirculation Fan Coolers, and Containment Spray,'' by incorporating
allowed outage times similar to those contained in NUREG 1431, Revision
0, ``Westinghouse Owner's Group Improved Standard Technical
Specifications,'' and by clarifying the operability requirements for
the service water pumps. The proposed changes would also clarify the
completion times for placing a unit in hot or cold shutdown if a
limiting condition for operation cannot be met.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
In accordance with the requirements of 10 CFR 50.91(a),
Wisconsin Electric Power Company (Licensee) has evaluated the
proposed changes against the standards of 10 CFR 50.92 and has
determined that the operation of Point Beach Nuclear Plant, Units 1
and 2, in accordance with the proposed amendments, does not present
a significant hazards consideration.
A proposed facility operating license amendment does not present
a significant hazards consideration if operation of the facility in
accordance with the proposed amendment will not:
1. Create 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. Will not create a significant reduction in a margin of
safety.
CRITERION 1
Operation of this facility under the proposed Technical
Specifications change will not create a significant increase in the
probability or consequences of an accident previously evaluated. The
proposed changes to the allowed out of service times have no impact
on the probability of an accident occurring. This equipment being
out of service is not an initiator for any accident previously
evaluated. There is no physical change to the facility, its systems
or its operation.
The clarification of service water pump operability requirements
will ensure redundant train capability to mitigate the consequences
of an accident which has been previously evaluated. Extending the
allowed out of service times for the safety injection, residual heat
removal, and containment spray pumps and valves and residual heat
removal heat exchangers does not create a significant increase in
the consequences of an accident previously evaluated. The proposed
changes are consistent with the Westinghouse Improved Standard
Technical Specifications, NUREG 1431, Revision 0. Plant specific
analysis demonstrates the proposed changes do not pose an undue risk
and thus will not result in a significant increase in the
consequences of an accident.
CRITERION 2
Operation of this facility under the proposed Technical
Specifications change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The safety injection, containment spray, and residual heat removal
pumps and valves and residual heat removal heat exchangers are used
to mitigate the consequences of an accident and are not normally in
use during power operation. The availability of these components
does not effect the possibility of a new or different type of
accident. The service water pumps are normally in use during power
operation. The proposed change will ensure that redundant train
capability exists. Minimum service water pump requirements remain
the same. The failure modes of the service water system remain
unchanged. Therefore, extending the allowed out of service time does
not create the possibility of a different type of accident than
previously evaluated.
CRITERION 3
Operation of this facility under the proposed Technical
Specifications change will not create a significant reduction in a
margin of safety. The proposed Technical Specification changes
revise the allowed outage times for the safety injection, residual
heat removal, and containment spray pumps and valves and residual
heat removal heat exchangers to 72 hours. This change will allow
more time for corrective maintenance to be performed on these
components, if required, and avoid potential transients and
challenges to safety systems associated with a required shutdown of
the unit without the specific safety related equipment operable. The
proposed changes are consistent with the Westinghouse Improved
Standard Technical Specifications, NUREG 1431, Revision 0. Plant
specific analysis demonstrates the changes do not pose an undue risk
and thus will not result in a significant reduction in a margin of
safety. The clarification to the specification for service water
pump operability may increase the margin of safety by ensuring that
redundant train capability exists.
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: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cynthia A. Carpenter
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: July 15, 1994
Description of amendment request: This amendment would revise
Technical Specification 3/4.3.3, Table 4.3-3, ``Radiation Monitoring
Instrumentation For Plant Operations Surveillance,'' to change the
analog channel operational test (ACOT) interval from monthly to
quarterly for the following radiation monitors: (1) Containment
Atmosphere - Gaseous Radioactivity - High (GT-RE-31 and 32); (2)
Gaseous Radioactive - RCS Leakage Detection (GT-RE-31 and 32); (3)
Particulate Radioactivity - RCS Leakage Detection (GT-RE-31 and 32);
(4) Fuel Building Exhaust - Gaseous Radioactivity - High (GG-RE-27 and
28); (5) Criticality - High Radiation Level (SD-RE-37 and 38; SD-RE-35
and 36); (6) Control Room Air Intake - Gaseous Radioactivity - High
(GK-RE-04 and 05).
This proposed change is identified as a line-item improvement in
Section 5.14 of Generic Letter 93-05, ``Line-Item Technical
Specifications Improvements to Reduce Surveillance Requirements for
Testing During Power Operations.''
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 of
Consequences of an Accident Previously Evaluated
The probability of occurrence and the consequences of an
accident evaluated previously in the Updated Safety Analysis Report
(USAR) are not increased due to the proposed technical specification
change. Review of past ACOT history for the affected monitors
revealed that these monitors have experienced no calibration or
setpoint-related problems since the beginning of plant operation.
Increasing the ACOT frequency for these monitors will not adversely
affect system operability, and this change would reduce the
potential for instrument damage, thus effectively increasing system
reliability and availability. These radiation monitors are not
accident-initiating equipment, so increasing the surveillance
interval on these monitors will not affect the probability of any
accident previously evaluated. In addition, for the monitors listed
in TS Table 4.3-3, no credit is taken in the plant accident analyses
in Chapter 15 of the USAR for any automatic actuation function
generated as a result of a radiation monitor signal. On these bases
it is concluded that the probability and consequences of the
accidents previously evaluated in the USAR are not increased.
2. Create the Possibility of a New or Different Kind of Accident
from any Previously Evaluated
No new type of accident or malfunction will be created since the
radiation monitors are not accident-initiating equipment. The
proposed change merely increases the ACOT interval for the affected
radiation monitors, and does not change the method and manner of
plant operation. The safety design bases in the USAR have not been
altered. Thus, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
3. Involve a Significant Reduction in the Margin of Safety
The proposed changes do not change the plant configuration in a
way that introduces a new potential hazard to the plant and do not
involve a significant reduction in the margin of safety. The
proposed changes do not affect applicable safety analysis acceptance
criteria and will not affect system operating conditions. In
addition, plant operating experience has shown that these monitors
have not experienced calibration of setpoint-related failures since
the beginning of plant operation. Therefore, it is concluded that
the margin of safety, as described in the bases to any technical
specification, is not reduced.
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 locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: Theodore R. Quay
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: July 22, 1994
Description of amendment request: The proposed amendment revises
Technical Specification (TS) 6.2.2.g, 6.3.1.b, and 6.12.1.c to reflect
title changes in the Wolf Creek Nuclear Operating Corporation (WCNOC)
organization. The title Supervisor Operations in TS 6.2.2.g is being
changed to Superintendent Operations. The title Radiation Protection
Manager in TS 6.3.1.b and the title Manager Radiation Protection in TS
6.12.1.c are being changed to Superintendent Radiation Protection. The
title changes do not represent any changes in reporting relationships,
job responsibilities, or overall organizational changes. This request
supersedes a request for amendment dated April 19, 1994, which was
noticed on June 22, 1994 (59 FR 32239)
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. Proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
These changes involve administrative changes to the WCNOC
organization and to the position titles and as such have no effect
on plant equipment or the technical qualification of plant
personnel.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. This change is administrative in nature and does not
involve any change to the installed plant systems or the overall
operating philosophy of Wolf Creek Generating Station.
3. The proposed change does not involve a significant reduction
in a margin of safety. This change does not involve any changes in
overall organizational commitments. A position title change alone
does not reduce the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
NRC Project Director: Theodore R. Quay
Peviously Published Notices Of Consideration Of Issuance Of
Amendments ToFacility 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.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of amendment request: June 3, 1994
Description of amendment request: In a letter of August 13, 1993,
and as supplemented on September 15, 1993, September 16, 1993, December
17, 1993, January 19, 1994, February 11, 1994, and February 24, 1994,
Commonwealth Edison Company submitted requests for amendments for steam
generator (SG) tube sleeving in accordance with (1) Westinghouse and
(2) Babcock & Wilcox processes. By letter dated March 4, 1994, the NRC
granted the proposed sleeving methods contingent upon four conditions
which the licensee accepted in their letter of February 24, 1994.
Three of the four changes will be reflected in the plants'
Technical Specifications (TS). By letter dated June 3, 1994, the
licensee requested changes to TS 3.4.5 and 3.4.6.2 to include the three
conditions, which are:
1. Amend the Byron and Braidwood licenses to reflect a primary-to-
secondary leakage rate limit of 150 gallons per day (gpd) through any
one SG.
2. Amend the Byron and Braidwood licenses to reflect an inservice
inspection of a minimum of 20 percent of a random sample of the sleeves
for axial and circumferential indication at the end-of-cycle. In the
event that an imperfection of 40 percent or greater depth is detected,
an additional 20 percent (minimum) of the unsampled sleeves should be
inspected, and if an imperfection of 40 percent or greater depth is
detected in the second sample, all remaining sleeves should be
inspected.
3. Add a condition to the Byron and Braidwood licenses to conduct
additional corrosion testing to establish the design life for the
kinetically or laser welded sleeved tubes in the presence of a crevice.
Collectively, these conditions will enable the licensee to have:
1. Further assurance that the integrity of the SGs will be
maintained in the event of a main steam line break or under loss-of-
coolant accident (LOCA) conditions;
2. Increased monitoring of the SG tube sleeves for any degradation;
and
3. Increased confidence that SG sleeve integrity will be maintained
for extended operations.
Date of publication of individual notice in Federal Register:
October 12, 1994 (59 FR 51613)
Expiration date of individual notice: November 14, 1994
Local Public Document Room location: For Byron, the Byron Public
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee
Street, Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Date of amendment request: September 7, 1994, and September 17,
1994 (two letters)
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) to incorporate a 1.0 volt
steam generator tube interim plugging criteria (IPC) for Unit 1
beginning with Cycle 7, which has begun. This supplements the
information that was published in the Federal Register on August 31,
1994 (59 FR 45019).
Date of publication of individual notice in Federal Register:
September 23, 1994 (59 FR 48917)
Expiration date of individual notice: October 24, 1994
Local Public Document Room location: Byron Public Library, 109 N.
Franklin, P.O. Box 434, Byron, Illinois 61010.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: September 16, 1994
Brief description of amendment request: The application changes the
Technical Specifications pertaining to the extension of the snubber
functional testing interval and the increase in sample plan size.
Date of publication of individual notice in Federal Register:
September 30, 1994 (59 FR 50019)
Expiration date of individual notice: October 31, 1994
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Power Authority of the State of New York, Docket Nos. 50-286 and
50-333, Indian Point Nuclear Generating Unit No. 3, Westchester
County, New York, and James A. FitzPatrick Nuclear Power Plant,
Oswego County, New York
Date of amendments request: September 16, 1994
Brief description of amendments: The proposed amendments would
revise Section 6.0 (Administrative Controls) of the Technical
Specifications of both facilities to reflect, in part, licensee
management changes. Specifically, the title of Executive Vice
President-Nuclear Generation is being changed to Executive Vice
President and Chief Nuclear Officer and a new position, Vice President
Regulatory Affairs and Special Projects, which will report to the
Executive Vice President and Chief Nuclear Officer, is being
established. In addition, the list of Safety Review Committee (SRC)
members is being deleted and replaced with a description of SRC
membership requirements, including individual qualifications. Each SRC
member, including the alternates, will have to be approved by the
Executive Vice President and Chief Nuclear Officer.
Date of publication of individual notice in Federal Register :
September 30, 1994 (59 FR 50021)
Expiration date of individual notice: October 31, 1994
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601 and the Penfield
Library, State University of New York, Oswego, New York 13126.
Power Authority of the State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New
YorkDate of application for amendment: September 29, 1994
Brief description of amendment: The proposed amendment would revise
Section 4.4 of the Indian Point Nuclear Generating Unit No. 3 Power
Plant Technical Specifications. Specifically, TS 4.4.E.1 would be
revised to allow a one-time extension to the 30-month interval
requirement for leak rate testing of Residual Heat Removal (RHR)
containment isolation valves AC-732, AC-741, AC-MOV-743, AC-MOV-744,
and AC-MOV-1870. A one-time schedular exemption from plant specific
requirements associated with 10 CFR Part 50, Appendix J, Type C testing
(local leak rate test) for the above listed RHR containment isolation
valves will be processed separately. This one-time extension for leak
rate testing of the RHR valves would defer the leak rate testing until
the next refueling outage, when the RHR system can be removed from
service as required by current procedures.
Date of publication of individual notice in Federal Register:
October 5, 1994 (59 FR 50777)
Expiration date of individual notice: November 4, 1994
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit 2, Somervell County, Texas
Date of amendment request: September 19, 1994
Brief description of amendment request: The proposed amendment
would revise the technical specifications for Comanche Peak Steam
Electric Station Unit 2 to allow a one-time extension of emergency
diesel generator and related surveillance testing from 18 to 24 months.
Date of individual notice in Federal Register: September 30, 1994
(59 FR 50024)
Expiration date of individual notice: October 31, 1994
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P. O. Box
19497, Arlington, Texas 76019
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: September 19, 1994
Brief description of amendment request: The proposed amendment
would revise the 18-month surveillance requirements of the technical
specifications for certain emergency core cooling system, containment
system, and plant systems to eliminate the restriction that these
surveillances be performed during shutdown or during the refueling mode
or cold shutdown.Date of individual notice in Federal Register:
September 30, 1994 (59 FR 50022)
Expiration date of individual notice: October 31, 1994
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P. O. Box
19497, Arlington, Texas 76019
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 20555, and at the local public document
rooms for the particular facilities 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 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: February 18, 1994, as
supplemented June 20, 1994
Brief description of amendments: These amendments allow credit to
be taken for burnup of spent fuel assemblies in establishing storage
locations within the spent fuel storage pool. The current spent fuel
storage pool is configured to store fresh fuel assemblies with a
maximum radially average enrichment of 4.30 weight percent (w/o) U-235
in a two-out-of-four checkerboard array. These amendments allow for
three distinct storage regions. Region 1 allows storage of fresh fuel
assemblies with a maximum radially averaged enrichment equal to 4.30 w/
o U-235 in a checkerboard configuration. Region 2 allows storage of
spent fuel assemblies in a three-out-of-four configuration. Region 3
allows storage of spent fuel assemblies in every location (four-out-of-
four configuration).
Date of issuance: September 30, 1994Effective date: September 30,
1994
Amendment Nos.: 82, 69, and 54
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17593) The supplemental letter dated June 20, 1994, responded to a
staff request for additional information, was clarifying in nature, and
did not affect the staff's initial no significant hazards
determination.The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 30, 1994.No
significant hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: August 18, 1994
Brief description of amendments: These amendments revised Technical
Specification 6.9.1.10 to add the analytical method supplement entitled
``Calculative Methods for the CE Large Break LOCA Evaluation Model for
the Analysis of CE and W Designed NSSS,'' CENPD-132, Supplement 3-P-A,
dated June 1985. This TS contains the list of analytical methods used
to determine the Palo Verde Nuclear Generating Station core operating
limits. Additionally, the existing references to earlier versions of
CENPD-132, and the associated approval letters are deleted.
Date of issuance: October 7, 1994
Effective date: October 7, 1994
Amendment Nos.: 83, 70, and 55
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 6, 1994 (59
FR 46069) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 7, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of application for amendments: July 6, 1994
Brief description of amendments: The NRC previously approved the
application of steam generator tube sleeving technologies through the
reference of specific vendor technical reports. These amendments remove
specific vendor technical report references and replace them with
references to the generic reports.
Date of issuance: September 29, 1994
Effective date: September 29, 1994
Amendment Nos.: 64, 64, 55, and 54
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39582) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 29, 1994. No
significant hazards consideration comments received: No
Local Public Document Room location: For Byron, the Byron Public
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee
Street, Wilmington, Illinois 60481.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
PointNuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: August 11, 1994
Brief description of amendment: The amendment revises Technical
Specification Section 6.5.1, Station Nuclear Safety Committee (SNSC),
to change the designation of the SNSC Chairman and to clarify the
maximum number of alternate members allowed for quorum purposes.
Date of issuance: October 3, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 177
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 1994 (59 FR
45002) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 3, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Consumers Power Company, Docket No. 50-155, Big Rock Point Plant,
Charlevoix County, Michigan
Date of application for amendment: June 11, 1993, as supplemented
July 1, 1993, and August 11, 1994.
Brief description of amendment: The amendment add acceptance
criteria for the electric and diesel fire pumps based on Emergency Core
Cooling System performance requirements and removes a portion of the
fire protection requirements from the Technical Specifications.
Date of issuance: September 30, 1994
Effective date: September 30, 1994
Amendment No.: 114
Facility Operating License No. DPR-6. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 7, 1993 (58 FR
36432). The July 1, 1993, and August 11, 1994, letters provided
clarifying information within the scope of the initial notice and did
not affect the staff's proposed no significant hazards considerations
findings. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 30, 1994.No
significant hazards consideration comments received: No.
Local Public Document Room location: North Central Michigan
College, 1515 Howard Street, Petoskey, Michigan 49770.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: May 24, 1994, as supplemented
August 4 and September 8, 1994
Brief description of amendments: The amendments transfer the boron
concentration in Technical Specification (TS) 3.9.1 for the reactor
coolant system and the refueling canal during MODE 6, and the boron
concentration in TS 4.7.13.3 for the spent fuel pool from the TS to the
Core Operating Limits Report (COLR). The associated Bases to the TS are
also changed. The application is submitted in response to the guidance
in Generic Letter 88-16 which addresses the transfer of fuel cycle-
specific parameter limits from the TS to the COLR.
Date of issuance: October 7, 1994
Effective date: To be implemented within 30 days from the date of
issuance
Amendment Nos.: 125 and 119
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 31, 1994 (59 FR
45022) The August 4 and September 8, 1994 supplemental submittals
provided clarifying information which did not affect the initial no
significant hazards determination. The Commission's related evaluation
of the amendments is contained in a Safety Evaluation dated October 7,
1994. No significant hazards consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: May 24, 1994 as supplemented
August 4 and September 8, 1994.
Brief description of amendments: The amendments transfer the boron
concentration values in TS 3.9.1 for the reactor coolant system and the
refueling canal during MODE 6, and the boron concentration value in TS
3/4.9.12 for the spent fuel pool from the TS to the Core Operating
Limits Report (COLR). The application is submitted in response to the
guidance in Generic Letter 88-16 which addresses the transfer of fuel
cycle-specific parameter limits from the TS to the COLR.
Date of issuance: October 12, 1994
Effective date: October 12, 1994
Amendment Nos.: 149 and 131
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 22, 1994, 59 FR
32228 The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 12, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: May 12, 1994, as supplemented
September 2, 1994
Brief description of amendment: The amendment revises Technical
Specification Sections 3.1 and 4.1 for Protective Instrumentation, the
associated bases, and tables to increase the surveillance test
intervals and add allowable out-of service times. The Technical
Specification changes will permit specified Channel Tests to be
conducted quarterly rather than weekly or monthly. The amendment will
enhance operational safety by reducing (1) the potential for
inadvertent plant scrams, (2) excessive test cycles or equipment, and
(3) the diversion of plant personnel and resources on unnecessary
testing.
Two additional technical changes have been incorporated. The fist
change involves extending the Channel Calibration interval for Average
Power Range Monitor. The second change would add a quarterly Channel
Calibration requirement for High Drywell Pressure (for Core Cooling)
and Turbine Trip Scram Instrumentation.
Editorial changes have been incorporated in Instrumentation
Sections 3.1 and 4.1 to provide clarity and consistency.
Date of issuance: October 11, 1994
Effective date: As of the date of issuance to be implemented within
90 days.
Amendment No.: 171
Facility Operating License No. DPR-16. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 22, 1994 (59 FR
32228). The September 2, 1994, submittal provided additional clarifying
information that did not change the initial proposed no significant
hazards consideration determination.The Commission's related evaluation
of this amendment is contained in a Safety Evaluation dated October 11,
1994.No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Florida Power and Light Company, et al., Docket No. 50-335 St.
Lucie Plant, Unit No. 1, St. Lucie County, Florida
Date of application for amendments: February 22, 1994
Brief description of amendments: This amendment modifies the
minimum stored borated water inventory requirements for Operational
Modes 1 through 4 by revising Figure 3.1-1 and Limiting Condition for
Operation 3.1.2.8 of the unit Technical Specifications (TS). The
associated bases for TS 3/4.1.2 are also revised to reflect the
bounding borated water makeup volumes, as a function of boric acid
concentration, which define the proposed inventory requirements.
Date of issuance: October 7, 1994
Effective date: October 7, 1994
Amendment No.: 129
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14888) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 7, 1994No significant
hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of application for amendments: July 19, 1994
Brief description of amendments: The amendments relocate certain
cycle-specific parameter limits from the Technical Specifications to
the Core Operating Limits Report.
Date of issuance: October 12, 1994
Effective date: October 12, 1994
Amendment Nos. 167 and 161Facility Operating Licenses Nos. DPR-31
and DPR-41: Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39587) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 12, 1994No significant
hazards consideration comments received: No
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of application for amendment: June 18, 1993, as supplemented
on December 17, 1993, and May 5, 1994.
Brief description of amendment: The amendment would revise the
Technical Specifications (TS) by clarifying TS wording for the Low
Pressure Coolant Injection (LPCI) and Containment Spray modes of the
Residual Heat Removal (RHR) system to assure consistency with
requirements of DAEC Updated Safety Analysis Report.
Date of issuance: October 4, 1994
Effective date: date of issuance to be implemented within 90 days
of issuance.
Amendment No.: 200
Facility Operating License No. DPR-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37074). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 4, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S. E., Cedar Rapids, Iowa 52401.
North Atlantic Energy Service Corporation, Docket No. 50-443,
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire
Date of amendment request: February 26, 1993 (License Amendment
Request 93-01), as modified by letter dated March 11, 1994, and April
7, 1993 (License Amendment Request 93-02), as modified by letter dated
February 24, 1994.
Description of amendment request: This amendment revises the
Appendix A Technical Specifications relating to the operability
requirements for the primary component cooling water (PCCW) system, the
service water (SW) system, and the ultimate heat sink (UHS). The
amendment redefines the requirements for operable PCCW and SW systems
and combines the technical specification requirements for the SW system
and the UHS. The changes affect Technical Specification sections 3/4
7.3, 3/4.7.4, and 3/4.7.5.
Date of issuance: October 5, 1994
Effective date: October 5, 1994
Amendment No.: 32
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications.
Date of initial notices in Federal Register: April 28, 1993 (58 FR
25860) June 23, 1993 (58 FR 34082). North Atlantic's letters dated
March 11, 1994 and February 24, 1994, provide additional clarifying
information related to risk calculations but neither letter changes the
initial proposed no significant hazards consideration determinations.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated October 5, 1994.No significant hazards
consideration comments received: No.
Local Public Document Room location: Exeter Public Library, 47
Front Street, Exeter, NH 03833.
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: May 6, 1994, as supplemented
August 16, 1994.
Brief description of amendment: The amendment modifies the Limiting
Conditions for Operation (LCO) for the Millstone Unit 2 Technical
Specifications (TS) 3.8.2.3 and 3.8.2.4 and the Surveillance
Requirements of TS 4.8.2.3.2.c.3. These changes relate to the amperage
requirements and the charging capability of the DC distribution
systems.
Date of issuance: October 14, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 180
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 22, 1994 (59 FR
32232) The August 16, 1994, 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 October 14, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
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: July 1, 1994
Brief description of amendment: The amendment revises the Technical
Specifications (TS) associated with the sump recirculation actuation
signal. The changes will be implemented after the installation of four
auctioneered power supplies in the Engineering Safety Feature Actuation
System sensor cabinets.
Date of issuance: October 7, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 179
Facility Operating License No. DPR-65. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 17, 1994 (59 FR
42342). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 7, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: April 23, 1994, as supplemented
August 4, 1994
Brief description of amendments: The amendments modify the
requirement for individuals filling certain plant management positions
to hold a Senior Reactor Operator (SRO) license. The amendments require
that only the Superintendent - Operations or the Assistant
Superintendent - Operations hold an SRO license.
Date of issuance: September 30, 1994
Effective date: September 30, 1994Amendment Nos. 80 and 41
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 23, 1993 (58 FR
34086) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated September 30, 1994.No
significant hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Public Service Electric and Gas
Company,Delmarva Power and Light Company, and Atlantic City
Electric Company,Docket No. 50-277, Peach Bottom Atomic Power
Station,Unit No. 2, York County, Pennsylvania
Date of application for amendment: May 13, 1994, as supplemented by
letter dated August 28, 1994
Brief description of amendment: This amendment allows a one-time
schedular extension of the second Type A Containment Integrated Leakage
Rate Test 10-year service period and an extended interval between Type
A tests.
Date of issuance: September 30, 1994
Effective date: September 30, 1994
Amendment No.: 196
Facility Operating License No. DPR-44: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 22, 1994 (59 FR
32235) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 30, 1994.No
significant hazards consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Philadelphia Electric Company, Public Service Electric and Gas
Company,Delmarva Power and Light Company, and Atlantic City
Electric Company,Docket Nos. 50-277 and 50-278, Peach Bottom Atomic
Power Station,Unit Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: June 9, 1994, as supplemented
by letter dated September 23, 1994.
Brief description of amendments: These amendments revise the
Technical Specifications (TS) surveillance requirements for scram
insertion times. The changes make the TS similar to those described in
NUREG-1433, ``Standard Technical Specifications General Electric
Plants, BWR/4.''
Date of issuance: September 30, 1994
Effective date: September 30, 1994
Amendments Nos.: 197 and 200
Facility Operating License Nos. DPR-44 and DPR-56: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37080) The September 23, 1994 letter provided clarifying information
that deletes language specifying the location for scram time acceptance
criteria and did not change the initial proposed no significant hazards
consideration. The Commission's related evaluation of the amendments is
contained in a SafetyEvaluation dated September 30, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: November 17, 1993, as
supplemented August 9, 1994
Brief description of amendment: The amendment revised the Technical
Specifications (TSs) to incorporate an instrument calibration
``allowable value'' format instead of the previous ``setting limit''
format. Instrumentation requiring specific value changes in the TSs
included:
(1) The overpressure protection system (OPS) actuation curve (TS
Figure 3.1.A-3).
(2) The minimum refueling water storage tank (RWST) water volumes
and low level alarm settings (specified in TS Section 3.3.A). In
addition the RWST level indicating switch calibration frequency
(specified in TS Table 4.1-1) was changed from once every 18 months to
once every 6 months.
(3) The control room ammonia and chlorine toxic gas instrument
settings (specified in TS Section 3.3.H).
(4) The containment pressure high and high-high engineered safety
features instrument settings (specified in TS Table 3.5.1).(5)
The main steam flow engineered safety features instrument settings
(specified in TS Table 3.5.1).
In addition, the TS Bases for protective instrumentation limiting
safety system settings (specified in TS Section 2.3) were revised to
clarify the description on constants K through K6 which are used
in the overtemperature delta-temperature and overpower delta-
temperature settings.
Date of issuance: October 7, 1994
Effective date: As of the date of issuance to be implemented prior
to restart from the current outage.
Amendment No.: 154
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 22, 1993 (58
FR 67860) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 7, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: July 25, 1994
Brief description of amendment: The Technical Specifications
amendment revised Table 3.6-1 (Non-Automatic Containment Isolation
Valves Open Continuously or Intermittently for Plant Operation) and
Table 4.4-1 (Containment Isolation Valves) to delete valves SI-1833A
and B and add valves SI-MOV-1835A and B. The valves being deleted no
longer perform a containment isolation function as a result of a
modification which removed the boron injection tank. The valves being
added are needed for testing the safety injection pumps.
Date of issuance: October 5, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 152
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 17, 1994 (59 FR
42346) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 5, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: August 4, 1994
Brief description of amendment: The amendment revises the fuel oil
availability requirements for the Emergency Diesel Generators (EDGs)
from Section 3.7 of the Technical Specifications (TSs). This TS change
requires that 30,026 gallons of fuel oil be available onsite in
addition to the oil in the EDG storage tanks. Specification 3.7.F.4 is
also being changed to require a total of 7056 gallons of fuel in the
EDG fuel oil storage tanks. In addition, administrative changes will
remove the word ``available'' from the phrase ''... gallons of fuel
available...'' in Section 3.7.A.5 (for the individual storage tanks) to
avoid confusion regarding the amount of usable fuel in the tanks.
Date of issuance: October 7, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 153
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 1994 (59 FR
45031) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 7, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: August 4, 1994
Brief description of amendment: The amendment revises Sections 3.4
and 3.5 of the Technical Specifications (TSs). The TS Section 3.4
revision reduces the maximum allowable percent of rated power
associated with inoperable Main Steam Safety Valves (MSSVs). This
change modifies Table 3.4-1 and the associated basis such that the
maximum power level allowed for operation with inoperable MSSVs is
below the heat removing capability of the operable MSSVs. The TS
Section 3.5 revision corrects administrative errors in the action
statements associated with Items 2.a and 2.c of Table 3.5-4.
Additionally, the changes to Item 2.b of Table 3.5-3 and Item 2.b of
Table 3.5-4 clarify the action statements associated with inoperable
high containment pressure (Hi-Hi Level) instrumentation.
Date of issuance: October 3, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 151
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 1994 (59 FR
45031) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated October 3, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, OhioDate of
application for amendment: September 3, 1992, as supplemented on
August 22, 1994
Brief description of amendment: The amendment revises the Technical
Specifications to include the maximum allowable steam generator level
as a variable limit based on the plant's mode of operation for Modes 1-
4 and to include additional shutdown margin requirements in Mode 3. The
amount of main steam superheat, the status of the main feedwater pumps,
and the status of the Steam and Feedwater Rupture Control System were
considered in determining the appropriate limits for the maximum
allowable steam generator level.
Date of issuance: October 7, 1994
Effective date: October 7, 1994
Amendment No. 192
Facility Operating License No. NPF-3. This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 28, 1992 (57 FR
48830) The August 22, 1994, submittal, provided additional supplemental
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
October 7, 1994.No significant hazards consideration comments received:
No
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, OhioDate of
application for amendment: March 30, 1994
Brief description of amendment: Revise T.S. to increase the
required boration flowrate in the event the required shutdown margin is
not met; increase the applicable minimum boron concentration and/or
volume requirements; revise the applicable Action statements and
surveillance requirements, and propose several administrative and
editorial changes.
Date of issuance: September 29, 1994
Effective date: date of issuance, to be implemented within 90 days
Amendment No. 191
Facility Operating License No. NPF-3. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27067) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 29, 1994.No
significant hazards consideration comments received: No
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: November 15, 1993
Brief description of amendments: The amendments revise the Comanche
Peak Steam Electric Station Units 1 and 2 technical specifications by
increasing the maximum permitted power at which the post-refueling
power ascension reactor coolant system flow verification can be
performed.
Date of issuance: October 7, 1994
Effective date: October 7, 1994, to be implemented within 30 days
of issuance.
Amendment Nos.: Unit 1 - Amendment No. 30; Unit 2 - Amendment No.
15
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 13, 1994 (59 FR
17606) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 7, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: March 28, 1994
Brief description of amendments: The amendments revise the
technical specifications by deleting reference to a large break LOCA
analysis methodology that is no longer applicable, and adding reference
to an approved steamline break analysis methodology.
Date of issuance: October 5, 1994
Effective date: October 5, 1994, to be implemented within 30 days
of issuance.
Amendment Nos.: Unit 1 - Amendment No. 28; Unit 2 - Amendment No.
14
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37088) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 5, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: April 25, 1994
Brief description of amendments: The amendments revise the TS
Surveillance Requirement 4.8.1.1.2 to allow ``slow starts'' of the
emergency diesel generator (EDG) instead of ``fast starts'' during the
monthly surveillance. A ``fast start'' is still required to be
performed at least once every 184 days. These changes are expected to
improve EDG availability and reliability.
Date of issuance: October 6, 1994
Effective date: October 6, 1994, to be implemented within 30 days
of issuance.
Amendment Nos.: Unit 1 - Amendment No. 29; Unit 2 - Amendment No.
15
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39599) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 6, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019.
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: April 15, 1994
Brief description of amendments: The amendments modify the
pressure/temperature operating limitations during heatup and cooldown
and the Low Temperature Overpressure Protection System pressure
setpoints and enabling temperatures for Units 1 and 2. The proposed
changes include revised Limiting Conditions for Operation, Action
Statements, and Surveillance Requirements for the power-operated relief
valves and block valves to address the concerns discussed in NRC
Generic Letter 90-06. The proposed changes also include several
editorial/administrative changes.
Date of issuance: October 5, 1994
Effective date: October 5, 1994
Amendment Nos.: 189 and 170
Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27069) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated October 5, 1994No significant
hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
NuclearPower Plant, Kewaunee County, Wisconsin
Date of application for amendment: May 26, 1994
Brief description of amendment: The amendment revises the Kewaunee
Nuclear Power Plant (KNPP) Technical Specification (TS) Sections 2.3,
3.6, and 4.6, by correcting minor typographical errors and format
inconsistencies. These changes are being made as a part of the
licensee's ongoing effort to revise each section of the KNPP TS to
achieve a consistent format and to convert the entire document to Word
Perfect. In addition, changes to the basis for TS Sections 2.3, 3.6,
and 4.6 have been made.
Date of issuance: September 29, 1994
Effective date: date of issuance, to be implemented within 30 days
Amendment No.: 111
Facility Operating License No. DPR-43. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39601) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated September 29, 1994.No
significant hazards consideration comments received: No.
Local Public Document Room location: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendments: May 30, 1991, as supplemented
May 7, 1993, and April 28, 1994.
Brief description of amendments: These amendments revised Technical
Specifications 15.3.1.A.5 and 15.3.15, and Table 15.4.1-1 and 15.4.1-2.
The changes specified more stringent limiting conditions for operation
and surveillance requirements for pressurizer power-operated relief
valves and block valves. These changes were proposed to conform to the
NRC's plan for resolution of Generic Issue 70, ``Power-Operated Relief
Valve and Block Valve Reliability,'' and Generic Issue 94, ``Additional
Low-Temperature Overpressure Protection for Light Water Reactors,'' as
conveyed in Generic Letter 90-06. Other related changes were also made.
Date of issuance: September 30, 1994
Effective date: September 30, 1994, to be implemented within 90
days.
Amendment Nos.: 155 & 159
Facility Operating License Nos. DPR-24 and DPR-27. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 28, 1993 (58 FR
16233). The May 7, 1993, and April 28, 1994, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination.The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated September 30, 1994. No significant hazards
consideration comments received: No.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Notice Of Issuance Of Amendments To Facility Operating Licenses And
Final Determination Of No Significant Hazards Consideration And
Opportunity For A Hearing (Exigent Public Announcement Or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC 20555,
and at the local public document room for the particular facility
involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By November 25, 1995, 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 20555 and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC 20555, by the
above date. Where petitions are filed during the last 10 days of the
notice period, it is requested that the petitioner promptly so inform
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, 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 the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Arizona Public Service Company, et al., Docket No. STN 50-529, Palo
Verde Nuclear Generating Station, Unit No. 2, Maricopa County,
Arizona
Date of amendment for amendment: October 9, 1994, as supplemented
by letter dated October 12, 1994
Brief description of amendment: The proposed amendment would modify
Technical Specification (TS) 4.8.2.1.e, ``DC Sources - Operating'' to
specify that the provisions of TS 4.0.1 and 4.0.4 are not applicable to
the battery capacity requirements until entry into Mode 4 coming out of
the fifth refueling outage or upon any deep discharge cycle of the
battery. The amendment was requested on an emergency basis so that the
licensee could declare the Unit 2 batteries operable based upon the
current capacities of the batteries without having to satisfy the
surveillance requirement of TS 4.8.2.1.e. The licensee will thus be
able to change modes and start up from the current mid-cycle steam
generator inspection outage.
Date of issuance: October 13, 1994
Effective date: October 13, 1994
Amendment No.: 71
Facility Operating License No. NPF-51: The amendment revised the
Technical Specifications.Public comments requested as to proposed no
significant hazards consideration: No.The Commission's related
evaluation of the amendment, finding of emergency circumstances, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated October 13, 1994.
Local
Public Document Room location: Phoenix Public Library, 12 East
McDowell Road, 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: Theodore R. Quay
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County,North Carolina
Date of application for amendments: September 9, 1994
Brief description of amendments: The amendments change the
Technical Specifications (TS) to revise the frequency for verifying the
position of the drywell-suppression chamber vacuum breakers when a
valve position indicator is inoperable from at least once every 72
hours to at least once every 14 days.
Date of issuance: October 5, 1994
Effective date: October 5, 1994
Amendment Nos.: 172 and 203
Facility Operating License Nos. DPR-71 and DPR-62. Amendments
revise the Technical Specifications.Public comments requested as to
proposed no significant hazards consideration: Yes. (59 FR 47648 dated
September 16, 1994) That notice provided an opportunity to submit
comments on the Commission's proposed no significant hazards
consideration determination. No comments have been received. The notice
also provided for an opportunity to request for a hearing by October 3,
1994, but indicated that if the Commission makes a final no significant
hazards determination, any such hearing would take place after issuance
of the amendment. The Commission's related evaluation of the amendments
and final no significant hazards consideration determination are
contained in a Safety Evaluation dated October 5, 1994.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: Mr. Mark S. Calvert, Associate General
Counsel, Carolina Power & Light Company, Brunswick Steam Electric
Plant, P. O. Box 10429, Southport, North Carolina 28461
NRC Project Director: Michael L. Boyle
Gulf States Utilities Company, Cajun Electric Power Cooperative,
and Entergy Operations, Inc., Docket No. 50-458, River Bend
Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: September 12, 1994, as supplemented
September 30, 1994.
Brief description of amendment: The amendment revises technical
specification 3/4.2.2, ``APRM Setpoints,'' to permit operation in
accordance with the Boiling Water Reactor Owners' Group (BWROG)
guidelines on improved BWR thermal-hydraulic stability.
Date of issuance: October 7, 1994
Effective date: October 7, 1994
Amendment No.: 75
Facility Operating License No. NPF-47. The amendment revised the
Technical Specifications. Public comments requested to proposed no
significant hazards consideration: Yes, September 21, 1994 (59 FR
48456). The Commission's related evaluation of the amendment, and final
determination of no significant hazards consideration are contained in
a Safety Evaluation dated October 7, 1994.Attorney for the licensee:
Mark Wetterhahn, Esq., Winston & Strawn, 1400 L Street, NW.,
Washington, D.C. 20005
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803
NRC Project Director: William D. Beckner
Dated at Rockville, Maryland, this 19th day of October, 1994.
For The Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II Office of Nuclear Reactor
Regulation
[Doc. 94-26422 Filed 10-25-95; 8:45 am]
BILLING CODE 7590-01-F