[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