[Federal Register Volume 59, Number 158 (Wednesday, August 17, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-10817]


[[Page Unknown]]

[Federal Register: August 17, 1994]


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Part II





Nuclear Regulatory Commission





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Operating Licenses, Amendments; No Significant Hazards Considerations; 

Biweekly Notices
UNITED STATES 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 July 25, 1994, through August 5, 1994. The 
last biweekly notice was published on August 3, 1994.

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 September 16, 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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit 
Nos. 1, 2, and 3, Maricopa County, Arizona

    Date of amendment requests: May 4, 1994
    Description of amendment requests: These amendment requests would 
revise Limiting Condition for Operation (LCO) 3.4.8.3 and Surveillance 
Requirement 4.4.8.3.1, ``Overpressure Protection Systems.'' 
Specifically, the LCO and surveillance requirements are revised to 
clarify that both shutdown cooling system (SCS) suction line relief 
valves shall be OPERABLE and aligned to provide overpressure protection 
not only during reactor (RCS) cooldown or heatup evolutions, but also 
during any steady state temperature periods maintained in the course of 
RCS cooldown or heatup evolutions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis about the issue of no significant hazards 
consideration, which is presented below:
    Standard 1 -- Involve a significant increase in the probability 
or consequence of an accident previously evaluated.
    The proposed amendments provide further clarification of the 
Technical Specifications and represent an additional operating 
limitation. Incorporating the noted clarification will not change 
the bases or assumptions contained in the safety analysis for this 
system. The most limiting low-temperature overpressure protection 
(LTOP) transients, the starting of an idle reactor coolant pump 
(RCP) and the inadvertent actuation of two high pressure safety 
injection (HPSI) pumps into a solid RCS, are not affected by the 
proposed clarification. Therefore, the proposed amendments do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    Standard 2 -- Create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    Clarifying the applicability of the LCO's and surveillance for 
steady state periods achieved and maintained during either a heatup 
or cooldown evolution does not modify the design or operation of 
plant equipment. No new or different failure modes will be 
introduced by incorporating this clarification into the LCO and 
surveillance requirement. Therefore, the proposed amendments will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Standard 3 -- Involve a significant reduction in a margin of 
safety.
    The clarification will enhance LCO 3.4.8.3 and Surveillance 
Requirement 4.4.8.3.1 for heatup and cooldown evolutions by ensuring 
operators are aware of this applicability during periods of steady 
state conditions. This clarification does not involve a change to 
safety limits, setpoints, or design margins. As such, the proposed 
amendments will not involve a significant reduction in a margin of 
safety at PVNGS.
    The NRC staff has reviewed the licensees' analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004
    Attorney for licensees: 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

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit 
Nos. 1, 2, and 3, Maricopa County, Arizona

    Date of amendment requests: June 17, 1994
    Description of amendment requests: The proposed amendments would 
increase the minimum nitrogen accumulator pressure for the atmospheric 
dump valves (ADVs), as stated in the surveillance requirements of 
Technical Specification (TS) 3/4.7.1.6. The change to the Bases 
increases the minimum time the ADV accumulators must be operable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis about the issue of no significant hazards 
consideration, which is presented below:
    Standard 1 -- Involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed Technical Specification change in the nitrogen 
accumulator supply minimum pressure will not increase the 
probability or consequences of any accident previously analyzed. The 
nitrogen accumulator pressure is normally maintained between 650-680 
psig. Nitrogen pressure from the accumulator is reduced to 105 psig 
prior to use in the operation of the ADVs. The pressure reduction 
will remain the same with the higher minimum accumulator pressure.
    Standard 2 -- Create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    Increasing the nitrogen accumulator minimum pressure does not 
create any new or different accidents than those previously 
evaluated. The normal air supply (the Instrument Air System) to the 
ADV is maintained between 105 to 125 psig. Currently, nitrogen from 
the accumulator is reduced to 105 psig prior to use in the ADV. The 
increased minimum pressure in the accumulator will still be reduced 
to 105 psig prior to use in the ADV.
    Standard 3 -- Involve a significant reduction in a margin of 
safety.
    The limitation on maintaining the nitrogen accumulator at a 
certain pressure is to ensure that a sufficient volume of nitrogen 
is in the accumulator to operate the associated ADV. Maintaining a 
higher minimum pressure ensures that sufficient nitrogen will be 
available to maintain the unit at HOT STANDBY for four hours and an 
additional 9.3 hours to reach COLD SHUTDOWN under natural 
circulation conditions in the event of failure of the normal control 
air system. Therefore, the proposed change does not involve a 
reduction in a margin of safety.
    The NRC staff has reviewed the licensees' analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004
    Attorney for licensees: 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

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit 
Nos. 1, 2, and 3, Maricopa County, Arizona

    Date of amendment requests: July 12, 1994
    Description of amendment requests: The proposed amendment would 
enhance the PVNGS Technical Specifications (TS) by adding a limiting 
condition for operation (LCO) action statement to Entry VIII B of Table 
3.3-3, ``Engineered Safety Features Actuation System Instrumentation.'' 
The proposed action statement would enhance safe plant operation by 
requiring timely plant shutdown if more than one of the new solid state 
degraded voltage relays in either train of 4.16kV are inoperable or not 
energized.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis about the issue of no significant hazards 
consideration, which is presented below:
    Standard 1--Involve a significant increase in the probability or 
consequence of an accident previously evaluated:
    The proposed amendment will add an action statement to TS Table 
3.3-3 entry VIII B which would allow eight hours to effect repairs. 
This action statement would be entered if more than one of the 
required four degraded voltage relays on either 4.16 kv bus is 
inoperable or not energized. If the eight hour allowed outage time 
is not met, the unit is placed in Hot Standby within six hours and 
in Cold Shutdown within the next thirty hours. Technical 
Specification 3.8.3.1 currently allows eight hours to restore a 4.16 
kv bus in the event of a loss of power to that bus. The loss of 
degraded voltage relays on that bus does not impact plant nuclear 
safety any more than the loss of the bus itself. Furthermore, even 
with the loss of all four degraded voltage relays monitoring one 
4.16 kv bus (for example, due to a blown 125 vdc circuit fuse), the 
loss-of-voltage relays on that bus, and the degraded voltage relays, 
as well as the loss-of-voltage relays monitoring the other bus would 
be unaffected. None of the UFSAR chapter 15 accident analyses are 
affected by this proposed amendment. The existing TS requirements 
and those components to which they apply are not altered by this TS 
amendment. There are no changes to the maintenance, surveillance, 
and/or qualification of any component/function in Table 3.3-3. 
Therefore, the addition of this proposed eight hour action statement 
to Table 3.3-3 entry VIII B does not increase the probability of 
occurrence or the consequences of any previously evaluated accident.
    Standard 2--Create the possibility of a new or different kind of 
accident from any accident previously evaluated:
    The TS requirements and the components to which they apply are 
not altered by this amendment. The new solid state degraded voltage 
relays in each 4.16 kv bus were installed under the 10 CFR 50.59 
change process. APS [Arizona Public Service Company] determined that 
the installation created no unreviewed safety question. This 
amendment has no impact on plant maintenance, testing, shutdown 
equipment, or component qualification. Plant operational safety is 
enhanced by this amendment. Therefore, the possibility of a new or 
different kind of accident is not created by this amendment.
    Standard 3--Involve a significant reduction in a margin of 
safety:
    The TS does not alter existing TS requirements or those 
components to which they apply. More specifically, there is no 
impact on safe plant shutdown, maintenance, containment isolation 
capability, containment leakage rate, or the operability of safety 
related valves. Therefore, the addition of the proposed action 
statement to the TS will not involve reduction in a margin of safety 
for fission product release to the atmosphere.
    The NRC staff has reviewed the licensees' analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004
    Basis for proposed no significant hazards consideration 
determination:
    Attorney for licensees: 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

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: June 8, 1994
    Description of amendments request: The proposed amendment would 
revise the Calvert Cliffs Nuclear Power Plant (CCNPP) Units 1 and 2 
Technical Specification (TS) Section 4.7.1.2.c to extend the interval 
for three Auxiliary Feedwater (AFW) surveillance requirements from 18 
to 24 months. Specifically, TS Section 4.7.2.c.1 requires the 
verification of each automatic valve in the flowpath actuate to its 
correct position and each AFW pump automatically start upon receipt of 
each AFW actuation system test signal; TS Section 4.7.2.c.2 requires 
verification that the AFW system is capable of providing a minimum 300 
gallons per minute nominal flow to each leg. This request is one of a 
series of proposed license amendments that would eliminate the need for 
mid-cycle surveillance outages by extending 18-month frequency 
surveillances to every refueling outage (nominally each 24 months).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The Auxiliary Feedwater (AFW) System provides a safety-related 
source of feedwater to the steam generators to mitigate design basis 
accidents involving loss of Main Feedwater. Failure of the AFW 
System is not an initiator for any previously analyzed accident. 
Therefore, the proposed change does not involve an increase in the 
probability of an accident previously evaluated.
    A historical review of surveillance test results and system 
performance indicates that the AFW System is very reliable. In 
addition, monthly surveillances of the AFW System will continue to 
verify proper pump and valve operation. The AFW System reliability 
and monthly surveillances provide assurance that undetected system 
degradation will not occur between 24-month surveillances. 
Therefore, the AFW System will continue to perform its safety 
function and there will be no significant increase in the 
consequences of accidents. Therefore, the proposed Technical 
Specification changes do not increase 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?
    This requested revision to increase the interval for some AFW 
surveillances from 18 to 24 months does not involve a significant 
change in the design or operation of the plant. No hardware is being 
added to the plant as part of the proposed change. The proposed 
change will not introduce any new accident initiators. Therefore, 
this change would not create the possibility of a new or different 
type of accident from any accident previously evaluated.
    3. Does operation of the facility in accordance with the 
proposed amendment involve a significant reduction in a margin of 
safety?
    The AFW System provides a margin of safety by providing a 
safety-related alternate supply of feedwater to the steam generator 
for removal of decay heat and cooldown of the Reactor Coolant 
System. The proposed changes do not affect the operation or design 
of the AFW System. Monthly surveillances and historical data provide 
assurance that the reduction in surveillance frequency will not 
adversely affect our ability to detect degradation in the 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: Michael L. Boyle

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: April 29, 1994
    Description of amendment request: This amendment is an additional 
followup to the amendment request of May 29, 1992, published in the 
Federal Register on July 8, 1992 (57FR30242), which changed the 
Technical Specifications Section 1.0, Definitions, to accommodate a 24-
month fuel cycle and which proposed the extension of the test intervals 
for specific surveillance tests. This amendment proposes extending the 
surveillance intervals to 24 months for the following additional 
surveillance tests:(1) Calibrate and test channels for Auxiliary 
Feedwater (AFW) initiation on steam generator water level (low-low).(2) 
Test channels for Auxiliary Feedwater initiation on trip of main 
feedwater pumps.The licensee's amendment proposal of November 25, 1992, 
requested approval for extending the surveillance interval of the 
Auxiliary Feedwater System to accommodate a 24-month fuel cycle and the 
approved change was issued in License Amendment No. 166. Subsequently, 
the licensee determined that two additional surveillances associated 
with this system had not been identified in the November 25, 1992, 
request. This amendment proposal requests approval of the additional 
surveillances. The changes requested by the licensee are in accordance 
with Generic Letter 91-04, ``Changes in Technical Specification 
Surveillance Intervals to Accommodate a 24-Month Fuel Cycle.''
    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 does not involve a significant hazards 
consideration since:
    1. There is no significant increase in the probability or 
consequences of an accident.
    The test results over the last four refuelings confirmed system 
operability with only one failure. This failure would not have 
impaired the ability of the auxiliary feedwater system to perform 
its intended safety function. The auxiliary feedwater system is 
redundant and diverse. The failure in the turbine driven pump did 
not impact the motor driven pumps.
    Based on the historical test data, it is concluded that no 
significant increase in the probability or consequences of an 
accident would be incurred by extending the operating cycle due to 
an increased surveillance interval.
    2. The possibility of a new or different kind of accident from 
any previously analyzed has not been created.
    The failure noted from the past test data appears random in 
nature and would not have defeated the redundancy in design that 
exists in the AFW system. The AFW system would have been capable of 
performing its intended safety function and therefore a new or 
different kind of accident would not have been created.
    3. There has been no reduction in the margin of safety.
    Past historical data demonstrates that the AFW systems would 
perform their safety function for an extended operating cycle should 
the surveillance period be extended by several months.
    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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: Pao Tsin Kuo

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: June 16, 1994
    Description of amendment request: The proposed amendment would 
revise License Condition 2.K of the license issued August 24, 1981, to 
provide for compliance with the NRC-approved fire protection program as 
described in the Updated Final Safety Analysis Report and for making 
changes to the NRC-approved fire protection program; would delete fire 
protection Technical Specification (TS) Sections 3.13 and 4.14 which 
contain limiting conditions for operation and surveillance 
requirements, respectively, for the high-pressure water fire protection 
system, fire protection spray systems, penetration fire barriers, fire 
detection systems, fire hose stations and hydrants, and the cable 
spreading room halon system; would delete Section 6.2.2.f which 
contains fire brigade staffing requirements; would delete Section 6.4.2 
which contains fire brigade training requirements; would add Section 
6.5.1.6.1 to add fire protection program responsibilities to the 
Station Nuclear Committee; would add Section 6.8.1.e to require written 
procedures and administrative policies for the fire protection program; 
would delete Section 6.9.2.b which requires a Special Report for 
inoperable fire protection and detection equipment; and would make 
corresponding changes to the Table of Contents and List of Tables.
    Generic Letter (GL) 86-10, dated April 24, 1986, and GL 88-12, 
dated August 2, 1988, from the NRC provided guidance to licensees to 
request removal of the fire protection TS. The licensee's proposed 
amendment is in response to these GLs.
    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 Commission has provided guidance concerning the application 
of the standards for determining whether a ``Significant Hazards 
Consideration'' exists by providing certain examples in 51 FR 7744 
(dated March 6, 1986). Example (vii) of those involving no 
significant hazards considerations relates to ``a change to conform 
a license to changes in the regulations, where the license change 
results in very minor changes to facility operations clearly in 
keeping with the regulations.''
    In this case, NRC Generic Letters 86-10 and 88-12, although not 
regulations, provide pertinent guidance relative to the above 
described proposed changes and implementation of the NRC fire 
protection regulations of 10 CFR 50.48(a). Specifically, the generic 
letters allow licensees to delete fire protection related technical 
specifications, provided that administrative requirements are added 
to technical specifications and a license condition is provided that 
requires the implementation and maintenance in effect of the 
approved fire protection program. Further, the generic letters 
provide for inclusion of the fire protection program into the UFSAR 
[Updated Final Safety Analysis Report] and permits future changes to 
the fire protection program without prior NRC approval, all as 
provided by the license condition and in accordance with the 
provisions of 10 CFR 50.59. Therefore, since the actions required by 
the generic letters have been taken and conform the license to the 
current interpretation of NRC fire protection regulations as 
described in Generic Letters 86-10 and 88-12, with no changes to 
facility operations, these proposed changes are in accordance with 
Example (vii) above.
    In accordance with the requirements of 10 CFR 50.92, the 
proposed changes to the Technical Specifications are deemed not to 
involve any ``Significant Hazards Considerations'' because operation 
of Indian Point Unit No. 2 in accordance with these changes would 
not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The fire protection program requirements are not affected in 
that the function, operation or surveillance requirements for any 
fire protection system or component are not being altered. The 
proposed changes simply relocate these requirements from the 
Technical Specifications to the UFSAR, are administrative in nature, 
and do not affect any other current plant equipment or practices. 
Therefore, the conclusions of current accident analyses are not 
affected. Further, as permitted by the proposed License Condition 
2.K, changes in the NRC-approved fire protection program will 
require an evaluation per the criteria of 10 CFR 50.59 to determine 
that the proposed change will not involve an unreviewed safety 
question. Therefore, future changes to the fire protection program 
will be evaluated in accordance with appropriate criteria.
    (2) Create the possibility for a new or different kind of 
accident from any previously evaluated.
    The proposed changes introduce no new mode of plant operation, 
do not involve physical modification to any structure, system or 
component, do not affect the function, operation or surveillance 
requirements for any equipment necessary for safe operation or 
shutdown of the plant or of fire protection equipment which protects 
such equipment, and do not involve any changes to setpoints or 
operating parameters. The changes are administrative only and all 
existing fire protection requirements are maintained. Therefore, the 
changes can not result in an unanalyzed accident. Further, as 
permitted by the proposed License Condition 2.K, changes in the NRC-
approved fire protection program will require an evaluation per the 
criteria of 10 CFR 50.59 to determine that the proposed change will 
not involve an unreviewed safety question. Therefore, future changes 
to the fire protection program will be evaluated in accordance with 
appropriate criteria.
    (3) Involve a significant reduction in the margin of safety.
    The existing fire protection program operability and 
surveillance requirements are retained as they are contained in the 
FPPP [Fire Protection Program Plan], and compliance will continue 
through proposed License Condition 2.K. Therefore, no margins of 
safety established by design or verified by testing to ensure 
operability of fire protection systems or components are affected. 
Further, as permitted by the proposed License Condition 2.K, changes 
in the NRC-approved fire protection program will require an 
evaluation per the criteria of 10 CFR 50.59 to determine that the 
proposed change will not involve an unreviewed safety question. 
Therefore, future changes to the fire protection program will be 
evaluated in accordance with appropriate criteria.
    Based on the above discussion, since these proposed changes to 
the Indian Point Unit No. 2 Technical Specifications satisfy the 
criteria specified in 10 CFR 50.92, are similar to an example 
provided by the Commission of a change which involves ``No 
Significant Hazards Considerations'', and are not similar to any 
examples that involve a ``Significant Hazards Consideration'', Con 
Edison has determined that this amendment application does not 
involve any ``Significant Hazards Considerations.''
    The proposed Technical Specification changes have been reviewed 
by the Station Nuclear Safety Committee and the Con Edison Nuclear 
Facilities Safety Committee. Both committees concur that these 
proposed changes do not represent any ``Significant Hazards 
Considerations.''
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: Michael L. Boyle

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: July 8, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification Section 3.7, Auxiliary Electrical 
Systems to clarify offsite power availability requirements and to 
revise emergency diesel generator fuel oil availability requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Consistent with the requirements of 10 CFR 50.92, the enclosed 
application involves no significant hazards based on the following 
information:
    1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response:
    Neither the probability nor the consequence of an accident 
previously analyzed is increased due to the proposed changes. There 
are no changes on the existing offsite power supply configuration or 
on the existing diesel fuel oil supply system or inventory 
requirements. This proposed amendment will allow for three diesel 
operation when a fuel oil storage tank or transfer pump is 
unavailable. In the event of an accident at this time, the three 
diesel operation would allow for more than minimum safeguards to be 
available, with maximum safeguards available for the first part of 
the event.
    2) Does the proposed license amendment create the possibility of 
a new or different kind of accident from any previously evaluated?
    Response:
    The existing 138 kV and 13.8 kV offsite power reliability is 
maintained with this change. There is no impact on availability of 
the alternate AC system, the three gas turbines, with this change. 
This change is consistent with the original licensing basis that the 
AEC accepted for the diesel fuel oil supply system.
    3) Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response:
    The proposed amendment does not involve a significant reduction 
in the margin of safety. The proposed amendment maintains the 
reliability of the preferred 138 kV and 13.8 kV offsite power and is 
consistent with the original licensing basis for diesel fuel oil 
inventory.
    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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: Robert A. Capra

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley 
Power Station, Unit No. 1, Shippingport, Pennsylvania

    Date of amendment request: July 29, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) 3/4.4.5 and 3/4.4.6.2 and 
associated bases to allow the implementation of interim steam generator 
tube plugging criteria for the tube support elevations during cycle 11. 
The allowed primary-to-secondary operational leakage from any one steam 
generator is proposed to be reduced from 500 gallons per day (gpd) to 
150 gpd. The total allowed primary-to-secondary operational leakage 
from all steam generators would be reduced from one gallon per minute 
(1440 gpd) to 450 gpd.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Testing of model boiler specimens for free span tubing (no tube 
support plate [TSP] restraint) at room temperature conditions 
show[s] burst pressures in excess of 5000 psi for indications of 
outer diameter stress corrosion cracking with voltage measurement as 
high as 19 volts. Burst testing performed on intersections pulled 
from BVPS [Beaver Valley Power Station] with up to a 2.7 volt 
indication shows measured burst pressure in excess of 6600 psi at 
room temperature. Burst testing performed on pulled tubes from other 
plants with up to 7.5 volt indications show[s] burst pressures in 
excess of 6300 psi at room temperatures. Correcting for the effects 
of temperature on material properties and minimum strength levels 
(as the burst testing was done at room temperature), tube burst 
capability significantly exceeds the safety factor requirements of 
RG [Regulatory Guide] 1.121. As stated earlier, tube burst criteria 
are inherently satisfied during normal operating conditions due to 
the proximity of the TSP. Test data indicates that tube burst cannot 
occur within the TSP, even for tubes which have 100 percent through 
wall electric discharge machining (EDM) notches, 0.75 inch long, 
provided that the TSP is adjacent to the notched area. Since tube to 
TSP proximity precludes tube burst during normal operating 
conditions, use of the criteria must retain tube integrity 
characteristics which maintain a margin of safety of 1.43 times the 
bounding faulted condition steam line break (SLB) pressure 
differential. As previously stated, the RG 1.121 criterion requiring 
maintenance of a safety factor of 1.43 times the SLB pressure 
differential on tube burst is satisfied by 7/8 inch diameter tubing 
with bobbin coil indications with signal amplitudes less than 8.82 
volts regardless of the indicated depth measurement. The plugging 
criteria (resulting in a projected end-of-cycle [EOC] voltage) 
compares favorably with the 8.82 volt structural limit considering 
the extremely slow apparent voltage growth rate of indications at 
BVPS. Using the established methodology of RG 1.121, the structural 
limit is reduced by allowances for uncertainty and growth to develop 
a beginning-of-cycle (BOC) repair limit which should preclude 
indications at EOC conditions which exceed the structural limit. The 
non-destructive examination (NDE) uncertainty component is 20.5 
percent and is based on the EPRI [Electric Power Research Institute] 
Alternate Repair Criteria (ARC). A bounding growth allowance of 40 
percent will be applied. This value is conservative for BVPS Unit 1. 
The BOC maximum allowable repair limit should not permit the 
existence of EOC indications (when the 40 percent growth and 20.5 
percent uncertainty allowances are applied) which exceed the 8.82 
volt structural limit. By adding NDE uncertainty allowances and an 
allowance for crack growth to the repair limit, the structural limit 
can be validated. Therefore, the maximum allowable BOC repair limit 
(RL) based on the structural limit of 8.82 volts can be represented 
by the expression:
    RL + (0.205 X RL) + (0.40 x RL) = 8.82 volts, or the maximum 
allowable BOC repair limit can be expressed as:
    RL = 8.82 volt structural limit/1.605 = 5.5 volts.
    It is reasonable that this repair limit (5.5 volts) could be 
applied for IPC [interim plugging criterion] implementation to 
repair bobbin indications greater than 1.0 or 2.0 volts independent 
of RPC [rotating pancake coil] confirmation of the indication. The 
analyses were performed based on a 1.0 or 2.0 volt repair limit. 
Duquesne Light Company has chosen to use a steam generator tube 
repair limit of 1.0 volt. Conservatively, an upper limit of 3.6 
volts will be used to assess tube integrity for those bobbin 
indications which are above 1.0 volt but do not have confirming RPC 
calls. This 3.6 volt upper limit for non-confirmed RPC calls is 
consistent with other recently approved IPC programs for the two 
other plants with 7/8 inch tubing that currently implement IPCs. 
Since the upper bound for repair of non-confirmed RPC is limited to 
a value far less than the limit associated with a full alternate 
criteria, the establishment of the repair limits are [is] judged to 
be independent of the pulled tube data base used.
    The conservatism of the growth allowance used to develop the 
repair limit is shown by the most recent BVPS eddy current data. The 
average voltage growth for all indications was 16 percent while the 
average voltage growth for indications greater than 0.75 volts at 
BOC was 6 percent. The largest overall voltage growth in a 
particular steam generator was found in the ``A'' steam generator, 
which had an overall average growth of 25 percent. Only two tubes 
had an absolute voltage growth which exceeded 1.0 volt for Cycle 9. 
The maximum absolute voltage growth in the 1993 inspection was 
recorded to be 1.18 volts. Each of the last three inspections, which 
included 100 percent of all hot leg tubes, showed decreasing voltage 
growth trends in each successive inspection for all categories; 
overall voltage growth, growth of BOC indications less than 0.75 
volts, and growth of indications greater than 0.75 volts. The 
decreasing voltage growth rate trend data indicates that DLC has 
good control of the ODSCC [outer diameter stress corrosion cracking] 
occurring in the BVPS Unit 1 steam generators and also implies that 
atypical voltage growth of a few indications is unlikely.
    Relative to the expected leakage during accident condition 
loadings, it has been previously established that a postulated main 
SLB [steam line break] outside of containment but upstream of the 
main steam isolation valve (MSIV) represents the most limiting 
radiological condition relative to the IPC. In support of 
implementation of the interim plugging criteria, it will be 
determined whether the distribution of cracking indications at the 
TSP intersections at the end of Cycle 11 are [is] projected to be 
such that primary-to-secondary leakage would result in site boundary 
doses within a small fraction of the 10 CFR 100 guidelines. A 
separate calculation has determined this allowable SLB leakage limit 
to be 6.6 gpm in the faulted loop. This limit was calculated using 
the Technical Specification RCS [reactor coolant system] Iodine-131 
activity level of 1.0 micro Curies per gram dose equivalent Iodine-
131 and the recommended Iodine-131 transient spiking values 
consistent with NUREG-0800. The projected SLB leakage rate 
calculation methodology prescribed in Section 3.3 of draft NUREG-
1477 will be used to calculate EOC leakage. The log-logistic 
probability of leakage correlation will be used to establish the SLB 
leak rate used for comparison with the 6.6 gpm faulted loop 
allowable limit. Due to the relatively low voltage levels of 
indications at BVPS and low voltage growth rates, it is expected 
that the actual calculated leakage values will be far less than this 
limit. Additionally, the current Iodine-131 levels as of May 1994 at 
BVPS are about 1000 times less than the Technical Specification 
limit of 1.0.
    Application of the criteria requires the projection of 
postulated SLB leakage, based on the projected EOC voltage 
distribution for the upcoming cycle. Projected EOC voltage 
distribution is developed using the most recent EOC eddy current 
results and a voltage measurement uncertainty. Data indicate that a 
threshold voltage of 2.8 volts would result in through wall cracks 
long enough to leak at steam line break conditions. Draft NUREG-1477 
requires that all indications to which the IPC are applied must be 
included in the leakage projection. Tube pull results from another 
plant with 7/8 inch tubing with a substantial voltage growth data 
base have shown that tube wall degradation of greater than 40 
percent through wall was readily detectable either by the bobbin or 
RPC probe. The tube with maximum through wall penetration of 56 
percent (42 percent average) had a voltage of 2.02 volts. This 
indication also was the largest recorded bobbin voltage from the EOC 
eddy current data. Based on the BVPS pulled tube and industry pulled 
tube data supporting a lower threshold for SLB leakage of 2.8 volts, 
inclusion of all IPC intersections in the leakage model is quite 
conservative. The ODSCC occurring at BVPS has historically resulted 
in relatively low voltage levels and has exhibited decreasing 
voltage growth trends over the last three inspections. BVPS has not 
identified ODSCC as a contributor to operational leakage. The 
current leakage levels at BVPS are negligible (less than 1 gpd). In 
order to satisfy the requirements of draft NUREG-1477, EOC 10 eddy 
current data will be used to calculate the projected SLB leakage 
according to draft NUREG-1477 methodology. Leakage calculated using 
the recommended EPRI leakage correlation will also be provided. 
Duquesne Light Company is requesting that the NRC review and approve 
the EPRI SLB leakage calculation methodology. Sufficient 
justification is included to establish acceptability of the EPRI 
leakage correlation based on criteria provided by the NRC in the 
February 8, 1994, Industry/NRC working meeting on the voltage based 
criteria.
    In order to assess the sensitivity of application of the voltage 
based criteria upon SLB leakage, the EOC 9 eddy current results were 
used to calculate postulated EOC 10 leakage using both the NUREG-
1477 methodology and EPRI correlation assuming that a 1.0 or 2.0 
volt plugging limit were implemented at the BOC 10.
    Results indicate SLB leakage of 0.46 gpm and 0.044 gpm using the 
NUREG and EPRI methodologies with an assumed probability of 
detection (POD) of 0.6 for a 2.0 volt repair limit. Since Duquesne 
Light Company has chosen to limit the voltage based plugging limit 
at 1.0 volt, EOC 11 SLB leakage is analyzed to be approximately 5 
percent lower than the calculated SLB leakage with a 2.0 volt repair 
limit.
    Therefore, implementation of the interim plugging criteria does 
not adversely affect steam generator tube integrity and 
implementation will be shown to result in acceptable dose 
consequences, therefore, the proposed amendment does not result in 
any increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Implementation of the proposed steam generator tube interim TSP 
plugging criteria does not introduce any significant changes to the 
plant design basis. Use of the criteria does not provide a mechanism 
which could result in an accident outside of the region of the TSP 
elevations; no ODSCC that has been identified at the TSP has been 
detected outside the thickness of the TSPs. Neither a single or 
multiple tube rupture event would be expected in a steam generator 
in which the plugging criteria has been applied (during all plant 
conditions).
    Specifically, Duquesne Light Company will implement a maximum 
leakage rate limit of 150 gpd per steam generator to help preclude 
the potential for excessive leakage during all plant conditions. The 
technical specification limits on primary-to-secondary leakage at 
operating conditions are to be a maximum of 450 gpd for all steam 
generators, or, a maximum of 150 gpd for any one steam generator. 
The RG 1.121 criterion for establishing operational leakage rate 
limits that require plant shutdown are based upon leak-before-break 
considerations to detect a free span crack before potential tube 
rupture during faulted plant conditions. The 150 gpd limit should 
provide for leakage detection and plant shutdown in the event of the 
occurrence of an unexpected single crack resulting in leakage that 
is associated with the longest permissible crack length. RG 1.121 
acceptance criteria for establishing operating leakage limits are 
based on leak-before-break considerations such that plant shutdown 
is initiated if the leakage associated with the longest permissible 
crack is exceeded.
    The single through wall crack lengths that result in tube burst 
at 1.43 times the steam line break pressure differential and SLB 
pressure differential alone are approximately 0.57 inch and 0.84 
inch, respectively. A leak rate of 150 gpd will provide for 
detection of 0.41 inch long cracks at nominal leak rates and 0.62 
inch long cracks at the lower 95 percent confidence level leak 
rates. Since tube burst is precluded during normal operation due to 
the proximity of the TSP to the tube and the potential exists for 
the crevice to become uncovered during SLB conditions, the leakage 
from the maximum permissible crack must preclude tube burst at SLB 
conditions. Thus, the 150 gpd limit provides for plant shutdown 
prior to reaching critical crack lengths for SLB conditions using 
the lower 95 percent leakage data. Additionally, this leak-before-
break evaluation assumes that the entire crevice area is uncovered 
during blowdown. Partial uncovery will provide benefit to the burst 
capacity of the intersection. Analyses have shown that only a small 
percentage of the TSPs are deflected greater than the TSP thickness 
during a postulated SLB.
    Steam generator tube integrity continues to be maintained 
through inservice inspection and primary-to-secondary leakage 
monitoring, therefore, the possibility of a new or different kind of 
accident from any accident previously developed is not created.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The use of the voltage based bobbin probe interim TSP elevation 
plugging criteria is demonstrated to maintain steam generator tube 
integrity commensurate with the requirements of RG 1.121. RG 1.121 
describes a method acceptable to the NRC staff for meeting GDCs 14, 
15, 31, and 32 by reducing the probability or the consequences of 
steam generator tube rupture. This is accomplished by determining 
the limiting conditions of degradation of steam generator tubing, as 
established by inservice inspection, for which tubes with 
unacceptable cracking should be removed from service. Upon 
implementation of the criteria, even under the worst case 
conditions, the occurrence of ODSCC at the TSP elevations is not 
expected to lead to a steam generator tube rupture event during 
normal or faulted plant conditions. The EOC distribution of crack 
indications at the TSP elevations will be confirmed to result in 
acceptable primary-to-secondary leakage during all plant conditions 
and that radiological consequences are not adversely impacted.
    In addressing the combined effects of loss of coolant accident 
(LOCA) and safe shutdown earthquake (SSE) on the steam generator 
component (as required by GDC 2), it has been determined that tube 
collapse may occur in the steam generators at some plants. This is 
the case as the TSP may become deformed as a result oflateral loads 
at the wedge supports at the periphery of the plate due to the 
combined effects of the LOCA rarefaction wave and SSE loadings. 
Then, the resulting pressure differential on the deformed tubes may 
cause some of the tubes to collapse.
    There are two issues associated with steam generator tube 
collapse. First, the collapse of steam generator tubing reduces the 
RCS flow area through the tubes. The reduction in flow area 
increases the resistance to flow of steam from the core during a 
LOCA which, in turn, may potentially increase peak clad temperature 
(PCT). Second, there is a potential that partial through wall cracks 
in tubes could progress to through wall cracks during tube 
deformation or collapse.
    Consequently, since the leak-before-break methodology is 
applicable to the BVPS reactor coolant loop piping, the probability 
of breaks in the primary loop piping is sufficiently low that they 
need not be considered in the structural design of the plant. The 
limiting LOCA event becomes either the accumulator line break or the 
pressurizer surge line break. LOCA loads for the primary pipe breaks 
were used to bound the conditions at BVPS for smaller breaks. The 
results of the analysis using the larger break inputs show that the 
LOCA loads were found to be of insufficient magnitude to result in 
steam generator tube collapse or significant deformation. The LOCA 
and SSE tube collapse evaluation performed for another plant with 
Series 51 steam generators using bounding input conditions (large 
break loadings) is considered applicable to BVPS.
    Addressing RG 1.83 considerations, implementation of the bobbin 
probe voltage based interim tube plugging criteria is supplemented 
by: enhanced eddy current inspection guidelines to provide 
consistency in voltage normalization, a 100 percent eddy current 
inspection sample size at the TSP elevations, and RPC inspection 
requirements for the larger indications left inservice to 
characterize the principal degradation as ODSCC.
    As noted previously, implementation of the TSP elevation 
plugging criteria will decrease the number of tubes which must be 
repaired. The installation of steam generator tube plugs reduces the 
RCS flow margin. Thus, the implementation of the alternate plugging 
criteria will maintain the margin of flow that would otherwise be 
reduced in the event of increased tube plugging.
    Based on the above, it is concluded that the proposed license 
amendment request does not result in a significant reduction in 
margin with respect to plant safety as defined in the Final Safety 
Analysis Report or any Bases of the Technical Specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location:  B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
    Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg, 
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., 
Washington, DC 20037.
    NRC Project Director: Walter R. Butler, Director

Entergy Operations, Inc., Docket Nos. 50-313, Arkansas Nuclear One, 
Unit 1, Pope County, Arkansas

    Date of amendment request: June 22, 1994
    Description of amendment request: The proposed amendment revises 
technical specifications (TSs) related to the emergency feedwater 
system (EFW). The proposed changes extend the allowable outage time 
when one EFW train is inoperable from 36 hours to 72 hours and adapt 
other EFW sections from the ``Restructured Standard Technical 
Specifications for B&W Plants'' to the Arkansas Nuclear One, Unit 1 
(ANO-1) format.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The Emergency Feedwater (EFW) system mitigates the consequences 
of any event with a loss of normal feedwater. This system is not the 
initiator of any previously analyzed accident, and therefore, 
changes to the specifications applicable to the EFW system present 
no significant increase in the probability of any previously 
evaluated accident.
    The changes that revise the required Actions and Allowable 
Outage Times associated with the EFW system have been evaluated for 
their effect on the Core Damage Frequency (CDF) previously 
calculated in the ANO-1 Probabilistic Risk Assessment (PRA). The new 
ANO-1 CDF values, incorporating the proposed AOT extension, are 
4.73E-05 (for the turbine-driven EFW pump) and 4.70E-05 (for the 
motor-driven pump). These values do not exceed the NRC Safety Goal 
of 1.0E-04 per reactor year, as stated in the Federal Register 
50FR32138. The delta CDF associated with these changes (6.16E-07 for 
the turbine-driven EFW pump and 3.04E-07 for the motor-driven EFW 
pump) have been evaluated with respect to criteria contained in 
SECY-91-270, dated August 27, 1991, and NUMARC 91-04, dated January 
1992, and fall within the category of events of low risk 
significance requiring no compensatory measures. This evaluation has 
shown the risk associated with the proposed changes to pose no undue 
risk to public health and safety, to be categorized as having low 
risk significance, and involve no significant increase in the 
consequences of an accident previously evaluated.
    The changes revising the Limiting Conditions for Operation 
result in more restrictive controls on the operability of the motor-
driven EFW pump. The previous specification required operability of 
both EFW pumps when the reactor was heated above 280 deg.F. The 
proposed change requires the operability of the motor-driven EFW 
pump whenever the unit is above the cold shutdown condition and any 
steam generator is relied upon for heat removal. With this change, 
the motor-driven EFW pump is now required to be operable in a 
condition not previously specified, constituting an additional 
requirement not previously specified. This change does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    The changes revising the Limiting Conditions for Operation also 
incorporate an Allowable Outage Time for the turbine-driven EFW pump 
steam supply valves which was not previously specified. The 7 day 
AOT is reasonable based on:
    1. The redundant steam supply (from the opposite steam 
generator) to the turbine-driven EFW pump is operable,
    2. The motor-driven EFW pump is operable, and
    3. The probability of an event occurring that would require the 
inoperable steam supply valve to actuate is relatively low.
    The changes to the surveillance specifications clarify the 
proper conditions required for the operability test of the turbine-
driven EFW pump, and revise the requirement for the verification of 
proper EFW flow path valve alignment. The change clarifying the test 
conditions is required to ensure a sufficient steam supply to the 
turbine-driven EFW pump to perform the test. During plant startup, 
from an RCS temperature of 280 deg.F to an RCS temperature of 
approximately 525 deg.F (corresponding to a steam generator pressure 
of approximately 830 psig) the turbine-driven EFW pump is classified 
as available until operability is proven by successful completion of 
the surveillance requirement. The proposed changes state that the 
EFW pumps and their associated flow paths shall be operable when the 
RCS is above the cold shutdown condition with any steam generator 
relied upon for heat removal (motor-driven EFW pump) and when RCS 
temperature is greater than or equal to 280 deg.F (turbine-driven 
EFW pump). This specification requires that the flow paths be 
properly aligned to maintain operability and is as restrictive as 
the current TS 4.8.1.c. The revised specification incorporates a new 
requirement to verify operator flexibility in determining the method 
of verification. Some methods that could be considered as fulfilling 
this requirement would include valve alignment checks, or a flow 
test verifying a level decrease in the `Q' condensate storage tank 
with a corresponding level increase in both steam generators. These 
changes result in no significant increase in the consequences of an 
accident previously evaluated.
    The other proposed changes included in this submittal, including 
the Bases changes, are considered to be administrative in nature and 
have no effect on the consequences of an accident previously 
evaluated. Relocation of the Emergency Feedwater Initiation and 
Control (EFIC) requirements from Section 3.4 to Section 3.5 places 
the requirements for this instrumentation system with the 
requirements for other instrumentation systems, resulting in greater 
consistency throughout the ANO-1 TS. Information in the Bases 
associated with the EFIC system has been corrected to reflect the 
actual plant condition and resolve a conflict with the ANO-1 Safety 
Analysis Report. The Bases changes add clarifying information to aid 
the operator in determining the applicability of the various EFW 
specifications.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed changes introduce no new mode of plant operation. 
The EFW system is not an event initiator. It functions to mitigate 
the consequences of any event with a loss of normal feedwater.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The changes proposed to the Limiting Conditions for Operation 
associated with the EFW system are more conservative than the 
current specification, thus resulting in an increase in the margin 
of safety. The proposed changes to the actions required when both of 
the EFW trains are inoperable and the auxiliary feedwater pump is 
unavailable no longer require an immediate plant runback, that is 
currently required, which could introduce a plant transient, thus 
resulting in an increase in the margin of safety.
    The changes revising the Limiting Conditions for Operations also 
incorporate and Allowable Outage Time for the turbine-driven EFW 
pump steam supply valves which was not previously specified. The 7 
day AOT is reasonable based on:
    1. The redundant steam supply (from the opposite steam 
generator) to the turbine-driven EFW pump is operable,
    2. The motor-driven EFW pump is operable, and
    3. The probability of an event occurring that would require the 
inoperable steam supply valve to actuate is relatively low.
    The changes to the surveillance specifications clarify the 
proper conditions required for the operability test of the turbine-
driven EFW pump, and revise the requirement for the verification of 
proper EFW flow path valve alignment. The change clarifying the test 
conditions is required to ensure a sufficient steam supply to the 
turbine-driven EFW pump to perform the test. During plant startup, 
from an RCS temperature of 280 deg.F to an RCS temperature of 
approximately 525 deg.F (corresponding to a steam generator pressure 
of approximately 830 psig) the turbine-driven EFW pump is classified 
as available until operability is proven by successful completion of 
the surveillance requirement. The proposed changes state that the 
EFW pumps and their associated flow paths shall be operable when the 
RCS is above the cold shutdown condition with any steam generator 
relied upon for heat removal (motor-driven EFW pump) and when RCS 
temperature is greater than or equal to 280 deg.F (turbine-driven 
EFW pump). This specification requires that the flow paths be 
properly aligned to maintain operability and is as restrictive as 
the current TS 4.8.1.c. The revised specification incorporates a new 
requirement to verify proper alignment prior to relying upon any 
steam generator for heat removal. This allows the operator 
flexibility in determining the method of verification. Some methods 
that could be considered as fulfilling this requirement would 
include manual valve alignment checks, or a flow test verifying a 
level decrease in the `Q' condensate storage tank with a 
corresponding level increase in both steam generators.
    This change does involve an incremental reduction in the margin 
of safety since the extension of the EFW Allowable Outage Time from 
36 hours to 72 hours does result in a slight increase in the Core 
Damage Frequency (CDF) as calculated in the ANO-1 Probabilistic Risk 
Assessment. The new ANO-1 CDF values, incorporating the proposed AOT 
extension, are 4.73E-05 (for the turbine-driven EFW pump) and 4.70E-
05 (for the motor-driven EFW pump). These values do not exceed the 
NRC Safety Goal of 1.0E-04 per reactor year, as stated in the 
Federal Register 50FR32138. The CDF associated with these changes 
(6.16E-07 for the turbine-driven EFW pump and 3.04E-07 for the 
motor-driven EFW pump) have been evaluated with respect to criteria 
contained in SECY-91-270, dated August 27, 1991, and NUMARC 91-04, 
dated January 1992, and fall within the category of events of low 
risk significance requiring no compensatory measures. This reduction 
is not considered significant in that the increase in CDF has been 
evaluated as posing no undue risk to the public health and safety 
and is categorized as having low risk significance.
    The other proposed changes included in this submittal, including 
the Bases changes, are considered to be administrative in nature. 
Relocation of the Emergency Feedwater Initiation and Control (EFIC) 
requirements from Section 3.4 to Section 3.5 places the requirements 
for this instrumentation system with the requirements for other 
instrumentation systems, resulting in greater consistency throughout 
the ANO-1 TS. Information in the Bases associated with the EFIC 
system has been corrected to reflect the actual plant condition and 
resolve a conflict with the ANO-1 Safety Analysis Report. The Bases 
changes add clarifying information to aid the operator in 
determining the applicability of the various EFW specifications.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, 
ArkansasNuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, 
Arkansas

    Date of amendment request: June 20, 1994
    Description of amendment request: The proposed amendments revise 
the administrative and control sections of the technical specifications 
(TSs) for Arkansas Nuclear One, Units 1 and 2. The proposed changes 
relocate controls associated with the ``Review and Audit'' functions 
from the TSs to the Quality Assurance Program and relocate requirements 
for the audit of emergency and security plans and implementing 
procedures from the TSs to the respective emergency and security plans.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The proposed changes do not affect reactor operations or 
accident analyses, have no radiological consequences, and are 
considered to be purely administrative in nature. All requirements 
relocated from the TSs have been evaluated with respect to the four 
criteria of the NRC Final Policy Statement On Technical 
Specifications Improvements'' as presented in SECY-93-067, and found 
to meet none of the criteria for inclusion in the TS.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed changes introduce no new mode of plant operation 
and do not affect the operability of safety-related equipment. All 
requirements relocated or deleted from the TSs have been evaluated 
with respect to the four criteria of the NRC ``Final Policy 
Statement On Technical Specifications Improvements'' as presented in 
SECY-93-067, and found to meet none of the criteria for inclusion in 
the TS.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    Existing TS operability and surveillance requirements are not 
reduced by the proposed change, thus no margins of safety are 
reduced. All requirements relocated or deleted from the TSs have 
been evaluated with respect to the four criteria of the NRC ``Final 
Policy Statement On Technical Specifications Improvements'' as 
presented in SECY-93-067, and found to meet none of the criteria for 
inclusion in the TS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

Entergy Operations Inc., Docket No. 50-382, Waterford Steam 
ElectricStation, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: February 9, 1993 as supplemented by 
letter dated July 22, 1994
    Description of amendment request: The proposed amendment would 
revise Section 3.0 and 4.0 of the Technical Specifications (TSs) 
consistent with the provision and intent of Generic Letter (GL) 87-09 
dated June 4, 1987.
    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:
    TS 3.0.4 prevents entry into an operational mode or other 
specified condition unless Limiting Conditions for Operations (LCOs) 
are met without reliance on Action Requirements. The intent of this 
TS is to ensure that a higher mode of operation is not entered when 
equipment is inoperable or when parameters exceed their specified 
limits.
    The proposed change clarifies TS 3.0.4 such that LCOs with 
Action Statements that permit continued operation for an unlimited 
period of time are exempt from the restrictions of TS 3.0.4. This 
provision is modified to require an additional plant safety review 
prior to implementing additional exceptions to 3.0.4 other than 
those currently stated in the individual specifications. This 
proposed change is consistent with existing NRC regulatory 
requirements for LCOs.
    The proposed change to TS 4.0.3 incorporates a 24-hour delay in 
implementing the Action Statements due to a missed surveillance 
requirement when the Action Statements provide a restoration time 
that is less than 24 hours. As reflected in GL 87-09, this change is 
justified in that it is overly conservative to assume that systems 
or components are immediately inoperable when a surveillance 
requirement has not been performed. The NRC concludes in Generic 
Letter 87-09 that a 24-hour time limit balances the risks associated 
with an allowance for completing the surveillance within this period 
against the risks associated with the potential for a plant upset 
and challenge to safety systems when the alternative is a shutdown 
to comply with Action Statements before the surveillance can be 
completed. The NRC further states that the potential for a plant 
upset and challenge to safety systems is increased if surveillances 
are performed during actions to initiate a shutdown to comply with 
Action Requirements.
    TS 4.0.4 has been modified to note that its provisions shall not 
prevent passage through or to operational modes as required to 
comply with Action Requirements. This change is consistent with the 
intent of the existing TS and represents a clarification.
    No previously analyzed accident scenario is changed by the 
proposed TS changes described above. Initiating conditions and 
assumptions remain as previously analyzed.
    Therefore, the proposed changes will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    The proposed change to TS 3.0.4 is administrative in nature. 
Entry into an operational mode or other specified condition will be 
allowed for those specifications not currently stating an exception 
to 3.0.4 when 1) the applicable LCOs Action Requirement permits 
continued operation for an unlimited period of time and 2) the PORC 
[plant operations review committee] has reviewed and approved the 
exception.
    The proposed change to TS 4.0.3 will allow continued operation 
for an additional 24-hours after discovery of a missed surveillance. 
As reflected in GL 87-09, missing a surveillance does not mean that 
a component or system is inoperable. In most cases, surveillances 
provide positive verification of operability.
    The proposed change to TS 4.0.4 will alleviate conflict within 
the TS. The change is necessary to allow the plant to proceed 
through or to required operational modes to comply with Action 
Statements even if applicable Surveillance Requirements may not have 
been performed.
    These changes do not affect the operation of the plant or the 
manner in which it is operated.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed change to TS 3.0.4 is administrative in nature and 
will have no impact on any margin of safety.
    The proposed change to TS 4.0.3 will allow up to 24-hours to 
perform a missed surveillance. In some cases this will eliminate the 
need for a plant shutdown. As reflected in GL 87-09, the overall 
effect is an increase in plant safety by avoiding unnecessary 
shutdowns and associated system transients due to missed 
surveillances.
    The proposed change to TS 4.0.4 will eliminate an internal 
conflict within the TS and allow the plant to proceed through or to 
required operational modes to comply with Action Statements even if 
applicable Surveillance Requirements for that mode may not have been 
performed.The NRC staff has previously evaluated these change in 
Generic Letter 87-09 and determined that the TS modifications will 
result in improved TS.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

Florida Power and Light Company, et al., Docket No. 50-335, St. 
Lucie Plant, Unit No. 1, St. Lucie County, Florida

    Date of amendment request: July 28, 1994
    Description of amendment request: The amendment will revise 
Technical Specifications (TS) 3/4.4.13 to incorporate Low Temperature 
Overpressure Protection (LTOP) requirements similar to those 
recommended by the NRC staff via Generic Letter 90-06. The proposed 
changes are in accordance with the resolution of Generic Issue 94 for 
St. Lucie Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Pursuant to 10 CFR 50.92, a determination may be made that a 
proposed license amendment involves no significant hazards 
consideration if 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. Each standard is discussed as 
follows:
    (1)Operation of the facility in accordance with the proposed 
amendment would not involve a signifiant increase in the probability 
or consequences of an accident previously evaluated.
    The changes proposed for St. Lucie Unit 1 Technical 
Specifications (TS) 3/4.4.13 are similar to those recommended by the 
NRC staff via Generic Letter 90-06 for Low Temperature Overpressure 
Protection (LTOP) systems. On the basis of technical studies 
performed for Generic Issue 94, the staff concluded that LTOP system 
unavailability is a contributor to the risk associated with 
overpressure transients during the shutdown modes of plant 
operation. Revisions to the actions required and the time for 
completion of such actions, in the event that one or more Power 
Operated Relief Valves (PORV) become inoperable, provide more rigor 
than the existing specifications and are designed to increase LTOP 
system availability. The administrative restrictions do not change 
the results of existing analyses performed to evaluate postulated 
accidents but will improve the availability of systems designed to 
mitigate pressure transients that could occur within the LTOP range. 
Therefore, operation of the facility in accordance with the proposed 
amendment will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    (2)Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment will not change the physical plant or the 
modes of operation defined in the facility license. The changes do 
not involve the addition of new equipment or the modification of 
existing equipment, nor do they alter the design of St. Lucie plant 
systems. Therefore, operation of the facility in accordance with the 
proposed amendment would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3)Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendment provides additional administrative 
restrictions for the operation of LTOP equipment. The applicability 
of Limiting Conditions of Operation (LCO) involving the PORVs will 
be extended to include Operational MODE 6 when the head is on the 
reactor vessel, and the rigor of required actions and action 
compleiton times in the event that one or more PORVs become 
inoperable will be increased. Consequently, the risk of low 
temperature operations will be reduced and safety during the 
shutdown modes of operation will be enhanced. Therefore, operation 
of the facility in accordance with the proposed amendment would not 
involve a significant reduction in a margin of safety.
    Based on the discussion presented above and on the supporting 
Evaluation of Proposed TS Changes, FPL has concluded that this 
proposed license amendment involves no significant hazards 
consideration.
    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: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: Harold F. Reis, Esquire, Newman and 
Holtzinger, 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: Victor M. McCree (Acting)

Northeast Nuclear Energy Company, et al., Docket No. 50-336, 
Millstone Nuclear Power Station, Unit No. 2, New London County, 
Connecticut

    Date of amendment request: July 1, 1994
    Description of amendment request: The proposed amendment to the 
Technical Specification (TS) would:1. Modify the facility by providing 
an auctioneered power supply for the engineered safety feature 
actuation system (ESFAS) sensor cabinets;2 Reinstate the 2-out-of-4 
sump recirculation system (SRAS) logic;3. Change Table 3.3 of the 
(Safety Feature Actuation System Instrumentation) by adding Manual main 
steam isolation (MSI) (Trip Buttons); by removing note (f) which 
describes the SRAS logic as a modified 2-out-of-4 logic; and by 
replacing Action Statement 4 with an Action Statement that allows 
operation with a second inoperable channel, provided both channels are 
placed in the bypassed condition. 4. Add to the TS new limiting 
conditions for operation and new surveillance requirements together 
with BASES (TS 3.3.2.2 and 4.3.2.2.1 and 4.3.2.2.2) for the ESFAS 
sensor cabinet power supply drawers.
    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 changes do not involve an SHC [significant 
hazards consideration] because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    SRAS Logic Modification
    Implementation of the auctioneered power supply for the sensor 
cabinets will permit the reinstatement of the original 2-out-of-4 
(six possible combinations) logic for SRAS initiation. The current 
logic only has four possible combinations. Changing the minimum 
number of SRAS channels required to be operable from four to three 
does not significantly reduce the available actuation combinations. 
Operation with one channel inoperable will still provide a 2-out-of-
3 logic (three possible trip combinations). With the current SRAS 
logic, operation with one channel in bypass does not meet the single 
failure criterion for proper SRAS operation. Amendment No. 168 
prevents that condition.
    Allowing continued operation with three operable channels is 
consistent with the original Millstone Unit No. 2 Technical 
Specifications (prior to Amendment No. 168).
    Note (f), which describes the current logic, will no longer 
apply after the auctioneering circuit is installed. This note is for 
information only and has no associated action or surveillance 
requirements. Therefore, removal of note (f) cannot affect either 
the probability or consequences of a postulated accident.
    In addition to the change in the minimum number of channels 
required to be operable, Action statement 4 will be revised to allow 
a limited period of two hours when a second channel may be placed in 
bypass for performance of surveillance testing. This is acceptable 
due to the installation of the auctioneering circuit and restoration 
of the full SRAS logic. Prior to the implementation of the short 
term modifications and Amendment No. 168, Action Statement 2 also 
applied to the SRAS. That Action Statement allows two hours of 
operation with two channels out of service. However, Action 
Statement 2 requires one of the two channels to be placed in the 
tripped position.
    Postulating a LOCA [loss-of-coolant accident] and an additional 
failure, while in an action statement that specifies a maximum 
allowed outage time, is beyond the design basis of Millstone Unit 
No. 2. However, with one SRAS channel in bypass and one in the 
tripped position, an additional failure (such as the loss of a DC 
vital bus) following the onset of a LOCA could result in a false 
SRAS signal.
    From an overall safety perspective, the potential consequences 
from a false SRAS at the onset of a LOCA are more severe than those 
from the failure to automatically generate an actuation signal. 
Proposed Action Statement 4 would require actuation of the remaining 
channel (following a LOCA and a loss of DC bus as a second failure) 
to initiate the SRAS. The existing operation procedures instruct the 
operator to ensure that the SRAS actuation occurs when the refueling 
water storage tank level decreases to a predetermined value. In the 
unlikely event that a LOCA occurred while a Action Statement 4 and 
no SRAS was generated at the appropriate time due to an additional 
failure which prevents one channel from tripping, the SRAS would be 
manually initiated by the operator.
    The amount of time that Millstone Unit No. 2 would operate under 
Action Statement 4 (with two SRAS channels in bypass) is 
approximately 6 hours per month. This is based on the requirement to 
conduct monthly channel functional tests for the three operable 
channels. The probability of a LOCA occurring during these 
surveillance, while in Action Statement 4, with a subsequent failure 
of the remaining 2-out-of-2 SRAS logic, is very low.
    Sensor Cabinet Auctioneering
    The proposed new Technical Specification 3.3.2.2, which 
establishes the requirements for the ESFAS sensor cabinets power 
supply drawers, permits 48 hours to restore an inoperable sensor 
cabinet power supply drawer to operable status. A power supply 
drawer renders it inoperable, or if either its normal or backup 
power is not available.
    Existing Technical Specification 3.8.2.1 contains an 8-hour 
action statement for restoring the power sources (VA-10, 20, 30, and 
40) if they become inoperable. The proposed 48-hour action statement 
for the power supply drawers is appropriate since the sensor cabinet 
would remain functional if either normal or alternate power was not 
available. However, a LOCA and an additional failure while in the 
action statement could result in a false SRAS, since two channels 
would supplied from a single DC power supply.
    Prior to Amendment No. 168, operation with an inoperable power 
supply drawer could continue indefinitely, provided the provisions 
of Technical Specification 3/4.3.2 were followed. Operation with a 
power supply inoperable for an indefinite period of time places all 
the signals associated with that sensor cabinet in the tripped 
condition. This creates a 1-out-of-4 tripped condition for SRAS. In 
this condition, the single failure required to be postulated could 
result in a false SRAS actuation.
    This 48-hour action statement is consistent with other action 
statements for ESFAS such as Action Statement 1 of Table 3.3-3. 
Also, this is consistent with the current wording of Action 
Statement 4 which allows 48 hours to restore an inoperable channel 
to operable while operating with the modified 2-out-of-4 logic.
    MSI Trip Button Addition
    The manual trip buttons provide a mechanism for the control room 
operator to initiate an MSI trip. The proposed Technical 
Specification change will require that a plant shutdown be initiated 
if either channel is out of service for more than 48 hours, and 
establishes a requirement for surveillance testing every refueling 
outage. Including the trip buttons in the Technical Specifications 
and establishing operation and surveillance requirements ensures 
their operability commensurate with their safety significance.
    Based on the above, the changes to Technical Specification 3/4.3 
do not increase the probability or consequence of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    SRAS Logic Modification
    Changing the number of channels required to be operable from 
four to three is acceptable since the original 2-out-of-4 logic will 
be restored. This change only affects the number and combinations of 
actuation channels necessary to initiate a SRAS. There is no change 
to the source or types of initiators, nor is there a change to the 
automatic response resulting from a SRAS.
    Note (f), which described the modified logic, will no longer 
apply after the auctioneering circuit is installed. This note is for 
information only and has no associated action or surveillance 
requirements. Therefore, removal of note (f) cannot create a new or 
different kind of accident.
    New Action Statement 4 restores the ability to operate for an 
indefinite period of time with one channel in bypass and for a 
limited period of time while two channels are out of service. The 
change from the original action statement to require that both 
channels be in bypass will prevent a false SRAS in the unlikely (and 
beyond design basis) event of a LOCA with an additional failure of a 
DC bus while in an LCO [limiting condition for operation].
    Sensor Cabinet Auctioneering
    The addition of a Technical Specification for the sensor cabinet 
power supply drawers does not create a potential for a new or 
different kind of accident. This new specification implements more 
restrictive operating requirements for the sensor cabinets. These 
are necessary to ensure that the sensor cabinets are energized from 
their primary power supply. The new specification does not affect 
the initiation of a SRAS signal nor the type of signal produced.
    The auctioneering modification does bring two vital AC 
facilities together via isolation devices. This introduces a 
potential for a new type of failure mechanism. As described in 
Attachment 1, adequate isolation ensures that a failure on one side 
of an isolation transformer does not adversely degrade the other 
side.
    MSI Trip Button Addition
    The manual trip buttons provide a mechanism for the control room 
operation to initiate an MSI trip. The Technical Specification 
change will require that a plant shutdown be initiated if either 
manual trip channel is out of service for more than 48 hours, and 
establishes a requirement for surveillance testing every refueling 
outage. The trip buttons were installed during the 1992 outage. 
Establishing operability requirements and surveillance frequency 
cannot create a new or different kind of accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    The net effect of the proposed modifications is to improve the 
reliability of the ESFAS and restore the design 2-out-of-4 logic for 
the SRAS. The proposed modifications improve the availability of the 
ESFAS, and do not affect the vital AC instrument panels.
    The Technical Specification changes establish controls for the 
used of the SRAS with the restored logic configuration. The 
combination of the auctioneering of the power supplies, the 
restoration of the 2-out-of-4 logic, and the revised Technical 
Specifications restores the margin of safety and operational 
flexibility originally designed for the sensor cabinets.
    Based on the above, there is no reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.
    Attorney for licensee: Gerald Garfield, Esquire, Day, Berry & 
Howard, City Place, Hartford, Connecticut 06103-3499.
    NRC Project Director: John F. Stolz

PECO Energy Company, Docket No. 50-353, Limerick Generating 
Station, Unit 2, Montgomery County, Pennsylvania

    Date of amendment request: June 30, 1994
    Description of amendment request: This amendment would remove 
certain remote shutdown system control valves and primary containment 
isolation valves from Technical Specifications Tables 3.3.7.4-1 and 
3.6.3-1 respectively, as a result of eliminating the steam condensing 
mode of the Residual Heat Removal 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 Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    These proposed changes will result in abandoning in place 
certain remote shutdown system control valves and removing from 
service and abandoning in place certain Primary Containment 
Isolation Valves (PCIVs) associated with the Residual Heat Removal 
(RHR) system steam condensing mode, and will remove the interface 
between the High Pressure Coolant Injection (HPCI) and RHR systems, 
therefore changing the primary containment pressure boundary.
    The RHR system steam condensing mode is a non-safety related 
function of the RHR system; however, the pressure and structural 
integrity of the associated piping and valves are safety-related. 
These proposed changes will not affect any components required to 
perform the safety-related function of the RHR or HPCI systems.
    The ability of the RHR or HPCI systems to respond to an accident 
will not be degraded. Only valves specifically dedicated for use for 
the RHR system steam condensing mode will be abandoned in-place, or 
removed from the plant. The valves' handswitches which are part of 
the remote shutdown panel (RSP) controls, will be physically removed 
from the RSP, since they will not perform any function (i.e., the 
associated valves will have the electrical power removed). The 
flanges and penetration caps that will become part of the primary 
containment boundary will be periodically tested for leakage as 
required by TS and 10CFR50, Appendix J. All piping and components 
that will remain operable will meet the original design 
requirements. The other modes of operation of the RHR system (e.g., 
Low Pressure Coolant Injection (LPCI), Shutdown [C]ooling (SDC)) 
will not be affected by these changes. Therefore, the proposed TS 
changes do not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    No new failure modes of RHR or HPCI systems are created by the 
proposed TS changes. The proposed changes will have no impact on the 
existing High Energy Line Break (HELB) analysis for Limerick 
Generating Station (LGS). All valves or piping removed and/or 
abandoned in place, are dedicated specifically for the RHR system 
steam condensing mode, and will not affect the operation of any 
components or piping required for other modes of operation of the 
RHR or HPCI systems. Therefore, the proposed TS changes do not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The steam condensing mode is a non-safety related function of 
the RHR system and, therefore, is not addressed in the TS. This mode 
will be physically separated from the other modes of operation of 
RHR and HPCI systems, and consequently, will not preclude them from 
performing their safety-related functions. The remote shutdown 
system control valves to be abandoned in place are not being used 
presently, and the proposed changes will not impact the safety 
operation of LGS Unit 2. The primary containment penetration caps, 
safety-related pipe caps and the flanges replacing the removed PCIVs 
will be designed, fabricated and installed in accordance with the 
original design requirements, i.e., American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section 
III, 1971 Edition with Addenda through Winter 1971. The added 
penetration caps and flanges will be capable of maintaining the 
primary containment pressure boundary and isolation capabilities 
that were required of the PCIVs and will be tested for leakage 
periodically, as required by TS and 10 CFR 50, Appendix J. 
Additionally, all piping and components that will remain operable 
will meet original design requirements. Therefore, the proposed TS 
changes do not involve a 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: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia, 
Pennsylvania 19101
    NRC Project Director: Charles L. Miller

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment requests: July 19, 1994
    Description of amendment request: This amendment will change the 
Technical Specification 3.1.5 for each unit for the standby liquid 
control system (SLCS) to remove the operability requirement for the 
SLCS while in Operational Condition 5 (refueling) with any control rod 
withdrawn, and to delete the 18-month system surveillance requirement 
(Surveillance Requirement 4.1.5.d.3).
    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. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed Technical Specification change to delete the 
operability requirement for the SLC System in OPCON 5* (OPERATIONAL 
CONDITION 5 with any control rod withdrawn) does not affect the 
probability or consequences of an accident previously evaluated. 
Design basis accident mitigation scenarios for SSES in OPCON 5 do 
not depend on, or require, SLC operability; therefore, the proposed 
change to delete SLC operability in OPCON 5* does not affect the 
probability or consequences of an accident previously evaluated.
    The proposed Technical Specification change to delete 
Surveillance Requirement 4.1.5.d.3, 18 month SLC heater operability 
check, does not affect the probability or consequences of an 
accident previously evaluated. Regarding the SLC heater function, 
the operability of the SLC system depends on maintaining the 
temperature of the sodium pentaborate solution above 70 deg.F to 
prevent the boric acid from precipitating out of solution. SLC 
heater 'A' is used to maintain tank temperature between 85 deg. F 
and 95 deg.F, thus ensuring that the boric acid remains in solution. 
The operability of the heater 'A' is verified through the daily 
performance of Technical Specification Surveillance Requirement 
4.1.5.a.1, which checks SLC solution temperature, and a control room 
alarm. Heater 'B' functions to raise SLC solution temperature prior 
to the mixing of SLC chemicals - the mixing of sodium pentaborate 
and water is an endothermic (heat consuming) reaction. The 
operability of heater 'B' is verified at the time when chemicals are 
added to the SLC tank, since a precondition for adding the chemicals 
is using heater 'B' to increase tank temperature to 100 deg.F. 
Heater 'B' does not function to maintain tank temperature during 
normal operation. Therefore, the proposed change does not impact 
Susquehanna's ability to maintain SLC solution temperature and thus 
does not increase the probability or consequences of an accident 
previously evaluated.
    2. This proposal does not create the possibility of a new or 
different kind of accident or [sic] from any accident previously 
evaluated.
    The proposed Technical Specification change to delete the 
operability requirement for the SLC System in OPCON 5* does not 
create the possibility of a new or different kind of accident or 
[sic] from any accident previously evaluated. The purpose of the SLC 
System is to provide backup capability for bringing the reactor from 
full power to a cold, Xenon-free shutdown, assuming that none of the 
withdrawn control rods can be inserted. This bases is consistent 
with the required operability of the SLC System in OPCONs 1 & 2. The 
proposed change does not affect the ability of SLC to meet its 
design basis. No credit is taken for SLC in OPCON 5 to mitigate the 
effects of reactivity transients, and the SLC system is not designed 
to terminate an inadvertent criticality event during core 
alterations (OPCON 5) with vessel water level at least 22 feet above 
top of vessel flange. Therefore, no new or different accident 
scenarios are created by the proposed change.
    The proposed Technical Specification change to delete 
Surveillance Requirement 4.1.5.d.3, 18 month SLC heater operability 
check, does not create the possibility of a new or different kind of 
accident or [sic] from any accident previously evaluated. The 
proposed change does not affect systems, structures, or components 
(SSCs) or the operation of these [SSCs]. The heating and heater 
control subsystems of the SLC system will continue to function as 
they were designed. The proposed change does not alter the heating 
limits or the method for maintaining SLC solution temperature. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident or [sic] from any accident 
previously evaluated.3. This change does not involve a significant 
reduction in a margin of safety.
    The proposed Technical Specification change to delete the 
operability requirement for the SLC System in OPCON 5* does not 
involve a significant reduction in a margin of safety. The potential 
for a decrease in the margin of safety, under this proposed change, 
would be associated with periods during OPCON 5* when the SLC system 
was not operable. Allowing the SLC system to be inoperable during 
OPCON 5* with the vessel level at least 22 feet above top of vessel 
flange, represents no reduction in the margin of safety since the 
SLC System is not designed to terminate an inadvertent criticality 
event with a greater volume of water in the reactor. Having the SLC 
system inoperable in OPCON 5* with reactor water levels at normal 
operating volumes, does not significantly reduce the margin of 
safety because of the number of other design and operating features 
which act to prevent inadvertent criticality events. Adequate 
shutdown margin is maintained through design and administrative 
controls; including, Shutdown Margin Demonstration, Technical 
Specification 3.1.1, defueling and refueling procedures, and 
refueling interlocks. In addition, the Reactor Protection System 
monitors for recriticality and actuates the Control Rod Scram 
function if a significant reactivity addition is sensed.
    The proposed Technical Specification change to delete 
Surveillance Requirement 4.1.5.d.3, 18 month SLC heater operability 
check, does not involve a significant reduction in a margin of 
safety. Adequate controls are in place, independent of the 18 month 
heater operability check, to ensure that the temperature of the 
sodium pentaborate solution is maintained above 70 deg. F. These 
controls include Surveillance Requirement 4.1.5.a.1, which checks 
SLC solution temperature daily, a control room alarm on low and high 
temperature, and the ambient temperature conditions in the SLC area 
which prevent rapid changes in SLC solution temperature. Operability 
of the 'B' heater is not needed to maintain SLC solution 
temperature, and the operability of this heater is verified at the 
time when chemicals are added to the SLC tank.
    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: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: Charles L. MillerPower Authority of the State 
of New York, Docket No. 50-333, James A. FitzPatrick Nuclear Power 
Plant, Oswego County, New York
    Date of amendment request: June 13, 1994
    Description of amendment request: The proposed amendment would 
modify the Facility Operating License by removing License Condition 
2.E. This condition applies to the construction cleanup, restoration, 
and maintenance of transmission lines. It incorporated into the 
Facility Operating License the requirements of U.S. Department of 
Interior publication ``Environmental Criteria for Electric Transmission 
Systems'' - 1970. The proposed amendment was requested to eliminate 
duplication of regulatory authority by government agencies of the same 
activity.
    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:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change will remove a license condition unrelated to 
nuclear safety. License condition 2.E incorporated into the 
Operating License the requirements of U.S. Department of Interior 
publication ``Environmental Criteria for Electric Transmission 
Systems'' - 1970. The goal of this standard is to ``safeguard 
aesthetic and environmental values within the constraints imposed by 
the current state of high-voltage transmission technology.'' License 
condition 2.E addresses the preservation of the environment and 
natural resources. Removing this condition from the Facility 
Operating License has no bearing on plant safety or the health and 
safety of the public considering its non-nuclear safety nature. The 
transmission line right-of-ways maintained by the Authority are 
subject to regulation by other State and Federal agencies. Removal 
of this license condition will not affect operation of safety 
related structures, systems or components nor affect the quality 
assurance program at the FitzPatrick plant. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    License condition 2.E of the James A. FitzPatrick Plant 
Operating License applies to the construction cleanup, restoration, 
and maintenance of transmission lines. The Authority's transmission 
lines are managed under guidelines based on the ``Generic 
Transmission Line Right-of-Way Management'' plan requirements. The 
requirements imposed by the plan on the FitzPatrick transmission 
line right-of-ways exceed those of the U.S. Department of Interior 
publication referenced in license condition 2.E in both scope and 
details. Therefore, implementing the proposed change will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. involve a significant reduction in a margin of safety.
    License condition 2.E of the James A. FitzPatrick Operating 
License applies to the construction cleanup, restoration, and 
maintenance of transmission lines. The requirements imposed by this 
license condition are unrelated to nuclear safety.
    Continued operation of the plant without Condition 2.E does not 
involve a significant reduction in any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Michael L. Boyle

Power Authority of the State of New York, Docket No. 50-333, James 
A. FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: July 21, 1994
    Description of amendment request: The proposed changes would modify 
paragraph 2.C.(3) of the Facility Operating License and relocate fire 
protection requirements from the Technical Specifications to an 
administrative procedure. These changes are based on the guidance 
contained in NRC Generic Letter 86-10, ``Implementation of Fire 
Protection Requirements,'' and Generic Letter 88-12, ``Removal of Fire 
Protection Requirements from Technical Specifications.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment will not involve a significant hazards 
consideration as defined in 10 CFR 50.92, because:
    (1) This change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated 
because no modifications, no changes to operating procedure 
requirements, no reduction in administrative controls and no 
reduction in equipment reliability are being made as a result of 
these changes. This proposed amendment relocates the fire protection 
LCOs [Limiting Conditions for Operation] and Surveillance 
Requirements from the Technical Specifications to an Administrative 
Procedure. No significant changes in content are being made to the 
Technical Specification requirements that are being relocated. 
Operating limitations will continue to be in effect, and required 
surveillances will continue to be performed in accordance with 
written procedures and instructions auditable by the NRC.
    Although future proposed changes to the fire protection program 
elements previously located in the Technical Specifications will no 
longer be controlled by 10 CFR 50.36, proposed changes to the Fire 
Protection requirements will be controlled by the License Condition 
and plant procedures. Programmatic controls will continue to assure 
that fire protection program changes do not reduce the effectiveness 
of the program to achieve and maintain safe shutdown in the event of 
a fire.
    (2) The possibility of an accident or malfunction of a different 
type than evaluated previously in the safety analysis report is not 
created because no reduction to the fire protection requirements, no 
modifications, no changes to operating procedure requirements, no 
reduction in administrative controls and no reduction in equipment 
reliability are being made as a result of these changes. 
Programmatic controls will continue to assure that fire protection 
program changes do not reduce the effectiveness of the program to 
achieve and maintain safe shutdown in the event of a fire.
    (3) This proposed amendment does not involve a reduction to the 
approved fire protection program or Fire Protection Technical 
Specification requirements because the Technical Specification fire 
protection requirements are being relocated, with no significant 
change in content, to an administrative procedure. Since there is no 
reduction in the requirements, no modifications, no changes to 
operating procedure requirements, no reduction in administrative 
controls and no reduction in equipment reliability are being made as 
a result of these changes, there is no reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location:  Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Pao Tsin Kuo

Power Authority of The State of New York, Docket No. 50-286, Indian 
PointNuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: July 25, 1994
    Description of amendment request: The licensee has requested an 
amendment to the Technical Specifications (TS) to revise 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(B) and add valves SI-MOV-
1835A(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.
    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:
    Consistent with the criteria of 10 CFR 50.92, the enclosed 
application is judged to involve no significant hazards based on the 
following information:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of any accident 
previously evaluated?
    Response:
    The proposed license amendment does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated. The change permits the removal of the two 
containment isolation valves on the Boron Injection Tank (BIT) 
bypass line. A previous amendment [Amendment No. 139, issued on 
October 15, 1993] to the Operating License removed the functional 
requirement for the BIT. Consequently, the function of the BIT 
bypass line to provide a Safety Injection [SI] pump test flow path 
has been rendered obsolete, permitting removal of the bypass line 
and associated valves. The bypass line will be cut and capped to 
assure containment integrity, therefore eliminating the need for 
containment isolation valves SI-1833A and SI-1833B. Opening the BIT 
outlet valve [SI-MOV-1835A or B] permits operability testing of the 
SI pumps, and is consistent with the current provision permitting 
opening of the BIT bypass valves. The changes do not impact the 
current operability and surveillance requirements for the Safety 
Injection System.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any previously 
evaluated?
    Response:
    The proposed license amendment does not create the possibility 
of a new or different kind of accident from any previously 
evaluated. The change proposes to eliminate two containment 
isolation valves on the BIT bypass line whose function has been 
rendered obsolete by a previous amendment to the Operating License. 
The bypass line will be cut and capped to assure containment 
integrity, therefore eliminating the need for these containment 
isolation valves. Intermittent opening of the BIT outlet valve is 
consistent with the current provision permitting opening of the BIT 
bypass valves, thereby allowing operability testing of the SI pumps. 
The changes do not impact the operability or surveillance 
requirements for the Safety Injection System.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    The proposed license amendment does not involve a significant 
reduction in a margin of safety for the following reasons. 
Currently, an orientation deficiency with the inboard BIT bypass 
isolation valve exposes its stem packing to the non-isolable side of 
the valve. The modification corrects this problem by removing both 
isolation valves and capping the pipes to assure integrity of the 
Containment and Safety Injection System. Additionally, removal of 
the isolation valves removes the potential for containment leakage 
resulting from valve degradation. Finally, removal of the BIT bypass 
line and its associated isolation valves does not inhibit the 
ability to test the SI pumps since a previous modification approved 
in an Amendment to the Operating License removed the functional 
requirement for the BIT. Consequently, the SI pumps may be flow 
tested with the BIT inservice, rendering obsolete the function of 
the BIT bypass line. Intermittent opening the BIT outlet valve is 
consistent with the current provision permitting opening of the BIT 
bypass valves, thereby allowing operability testing of the SI pumps.
    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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019.
    NRC Project Director: Pao Tsin Kuo

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
County, New Jersey

    Date of amendment request: June 29, 1994
    Description of amendment request: These proposed amendments would 
revise the Technical Specifications to increase the minimum volume of 
oil contained in the Diesel Fuel Oil Storage Tanks (DFOSTs) at the 
Salem Generating Station (SGS). It would also revise the Updated Final 
Safety Analysis Report (UFSAR) description of the fuel oil storage 
system capability.
    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) [This proposal does] not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Emergency Diesel Generator (EDG) fuel oil is used to support 
mitigation of design basis events involving loss of the preferred 
(offsite) source of A.C. power. Fuel oil storage capacity has no 
effect on the probability of any accident previously evaluated.
    Onsite fuel oil storage capability is designed to provide 
assurance of long term diesel operation to mitigate the consequences 
of a design basis accident. The proposed change would increase the 
minimum required volume in the Seismic Class I Diesel Fuel Oil 
Storage Tanks (DFOSTs), and would revise the Updated Final Safety 
Analysis Report (UFSAR), as part of an effort to reconstitute the 
basis for SGS fuel oil storage capacity. The DFOST inventory at the 
proposed minimum Technical Specification (TS) limit, combined with 
the emergency fill connection and Seismic Class III Fuel Oil Storage 
Tank and transfer capability, would continue to provide a long term 
onsite fuel oil supply to the EDGs. Operations and Emergency 
Preparedness procedures would facilitate the transfer of fuel oil, 
and procurement from offsite sources as a contingency measure. 
Therefore, the ability to provide a long term supply of fuel oil to 
the EDG's is maintained, and the proposed change would not result in 
any significant increase in consequences of an accident previously 
evaluated.
    (2) [This proposal does] not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change would increase the minimum DFOST level 
required by TS, and redefines the fuel oil storage and transfer 
systems' capability based on plant specific fuel oil consumption 
rate and EDG load profiles. These changes would not result in 
operation in any configuration prohibited by the present TS, and do 
not introduce the possibility of any new type of accident.
    (3) [This change does] not involve a significant reduction in a 
margin of safety.
    The EDG fuel oil storage and transfer capability would continue 
to support reliable, long term EDG operation, thereby maintaining an 
acceptable margin of safety relative to the ability of onsite A.C. 
power to support operation of equipment important to safety. The 
proposed changes do not involve a significant reduction in margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: Charles L. Miller

Tennessee Valley Authority, Docket Nos. 50-259 and 50-296, Browns 
Ferry Nuclear Plant, Units 1 and 3, Limestone County, Alabama

    Date of amendment request: March 31, 1994 (TS 319)
    Description of amendment request: The proposed amendment revises 
the setpoints for instrumentation used to isolate high energy line 
breaks in the high pressure coolant injection (HPCI) and reactor core 
isolation cooling (RCIC) systems. The proposed amendment also defines 
specific areas where steam line space temperatures are monitored.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes to the HPCI and RCIC steam line space 
isolation setpoints do not affect any precursor for any design basis 
events or operational transients analyzed in the Browns Ferry Final 
Safety Analysis Report. Therefore, the probability of an accident 
previously evaluated is not increased.
    The HPCI and RCIC steam line space high temperature isolations 
are provided to ensure automatic closure of each system's primary 
containment isolation valves for a HPCI or RCIC steam line break. 
The isolation occurs when a very small leak has occurred. If the 
small leak is allowed to continue without isolation, offsite dose 
limits may be reached. As a result of the environmental 
qualification program, the environmental responses of the reactor 
building to high energy line breaks were analyzed. TVA used computer 
modeling techniques to predict the temperature response of various 
reactor building zones to high energy line breaks. The results 
indicate that the setpoints for the HPCI and RCIC temperatures 
switches should be lowered. The lower setpoints assure the timely 
initiation of a closure signal to the primary containment isolation 
valves. Therefore, assuring the maximum allowable temperatures are 
not exceeded.
    The proposed change to the HPCI and RCIC steam line space 
isolation setpoints are in the conservative direction and provides 
the same or earlier detection and isolation of HPCI and RCIC steam 
line breaks.
    The proposed trip level settings are high enough to ensure that 
spurious trips do not occur from normal or transient system 
operation and low enough to ensure that line breaks are detected and 
isolated before design conditions are exceeded. Therefore, the 
proposed changes will not significantly increase the consequences of 
an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed change to the HPCI/RCIC steam line space high 
temperature isolations does not involve any modification to plant 
equipment or changes in operating procedures. No new failure modes 
are introduced. There is no effect on the function or operation of 
any other plant system. No new system interactions have been 
introduced by the change. The results of a break in the HPCI or RCIC 
steam lines remain as before. The HPCI or RCIC steam line area 
temperature switches will still detect a break due to an increase in 
area temperature and provide an initiation signal to close the 
system primary containment isolation valves to prevent reactor 
coolant loss. The proposed change will conservatively serve to 
detect and mitigate HPCI and RCIC line breaks more expeditiously.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change will not reduce the margin of safety. The 
proposed change ensure that HPCI and RCIC steam line breaks are 
isolated at the same or lower steam line area temperatures. Computer 
modeling techniques were utilized to predict the temperature 
response in various areas through which the HPCI and RCIC steam 
lines pass. The revised setpoints are established above the maximum 
expected normal room temperatures to avoid spurious actions due to 
ambient conditions and below the analytical limits to ensure timely 
pipe break detection and isolation. Substantial margin exist between 
the maximum temperature expected in each area and the minimum 
actuation temperature determined for each temperature switch. With 
the substantial margin between maximum temperatures for the areas 
and the minimum actuation temperature of the switches, the maximum 
temperatures cannot result in actuation of the switches. The design 
and function of the affected components has not been changed. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Mr. Frederick J. Hebdon

Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear 
Plant, Unit 2, Limestone County, Alabama

    Date of amendment request: May 11, 1994 (TS 347T)
    Description of amendment request: The proposed amendment extends 
the allowed outage time for the Browns Ferry Nuclear Plant (BFN) Unit 2 
250 volt DC (direct current) control power supplies from 5 to 45 days. 
The amendment is a temporary revision to the BFN Unit 2 Technical 
Specifications (TS) to permit replacement of batteries and other 
hardware.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change involves temporarily (one-year period) 
extending the 5-day AOT [allowed outage time] for the 250-volt 
shutdown board control power supplies to 45 days. As such, this 
change does not increase the probability of any accident previously 
analyzed.
    The 250-volt DC Power System is required to function to mitigate 
the consequences of design basis accidents. The loss of a single 
250-volt DC shutdown board control power supply will result in a 
loss of control power for the 480-volt and the 4160-volt shutdown 
board that it serves. Loss of control power results in loss of only 
those engineered safeguards supplied by its respective shutdown 
boards. Redundant safe shutdown equipment exists to mitigate the 
consequences of design basis accidents. As discussed in Final Safety 
Analysis Report (FSAR) subsection 8.6.4.3, a single failure of a 
shutdown board control power supply is acceptable.
    Loss of a single 250-volt plant DC power supply will not prevent 
Unit 2 safe shutdown. The 250-volt plant DC power supply system is 
designed so that any two out of the three power supplies carry the 
entire load needed for safe shutdown. As discussed in FSAR 
subsection 8.6.4.2 a single failure of a 250-volt plant DC power 
supply is acceptable.
    At no time will control power be unavailable to the shutdown 
boards during the system upgrades. The proposed change will only 
increase the time allowed to operate the plant while a 250-volt DC 
shutdown board control power supply is out of service.
    The proposed TS change allows an additional 40 days to perform 
system upgrades and results in a small increase in risk. This small 
increase in risk is associated with the probability and consequences 
of a 250-volt plant DC power supply malfunction while it is 
supplying shutdown board control power. The increase in risk 
associated with extending the AOT was analyzed in a Probabilistic 
Safety Assessment (PSA) and determined to be approximately 0.3 
percent. This small increase in risk is determined to be 
insignificant and well within the uncertainty bounds of the PSA.
    The proposed TS change does not change the function of any plant 
structure, system or component. The proposed change allows for 
improvements to the 250-volt DC shutdown board control power supply 
system. The improvements will increase the capability and 
reliability of the system. Qualified backup power will be utilized 
at all times during system modifications. Only one power supply will 
be out of service at a time during the modifications.
    The small increase in risk is more than offset by the increased 
capability, capacity, and reliability of the new power supplies. 
Therefore, the power supply modifications will result in a net 
overall safety benefit.
    [The licensee has also committed to implement compensatory 
measures while performing the power supply modifications. These 
measures provide additional confidence that potential accident 
consequences are not increased.]
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Extending the 5-day AOT for the 250-volt shutdown board control 
power supplies to 45 days does not create the possibility of a new 
or different kind of accident, nor does it increase the probability 
that an accident will occur. The AOT extension does not involve 
plant modifications that could create the possibility of a new or 
different kind of accident from any of those discussed in the FSAR.
    The 250-volt DC shutdown board control power supply 
modifications involve replacement of the existing components with 
more reliable, increased capacity equipment having the same 
functions as before.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed TS change involves a risk increase of approximately 
0.3 percent. TVA [the Tennessee Valley Authority, the licensee] 
considers this small increase to be insignificant. TVA also 
considers that the small increase in risk is offset by the benefits 
associated with replacing the control power supplies with new, 
upgraded equipment. Therefore, the proposed TS change does not 
involve a significant reduction in a margin of safety.
    [The licensee has also committed to implement compensatory 
measures while performing the power supply modifications. These 
measures provide additional capability to mitigate an accident, 
minimizing any effect on safety margin.]
    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Mr. Frederick J. Hebdon

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: July 18, 1994
    Description of amendment request: The proposed amendment would 
modify Point Beach Nuclear Plant Technical Specification (TS) 15.3.7, 
``Auxiliary Electrical Systems,'' by including an allowed outage time 
for one of the four connected station battery chargers and subsequent 
shutdown requirements. The basis for Section 15.3.7 would also be 
revised to support the above changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    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.
    The proposed amendment allows operation for up to two hours with 
one out of the four connected station battery chargers out of 
service. The 2-hour outage time is based on Regulatory Guide 1.93 
and reflects a reasonable time to assess plant status and either 
connect an operable battery charger to the affected DC bus or 
prepare to effect an orderly and safe shutdown of the operating 
unit(s). Since the batteries, chargers, and their associated vital 
instrument buses provide sufficient redundancy to assure the 
initiation of proper protective actions during degraded system 
conditions, operation of PBNP in accordance with these proposed 
amendments cannot create an increase in the probability or 
consequences of an accident previously evaluated, create a new or 
different kind of accident, or result in a significant reduction in 
a margin of safety. Therefore, the proposed changes do not present a 
significant hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: John N. Hannon

Previously Published Notices Of consideration Of Issuance Of 
Amendments To Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, And Opportunity For A Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Georgia Power Company, Docket No. 50-366, Edwin I. Hatch Nuclear 
Plant, Unit 2, Appling County, Georgia

    Date of amendment request: July 19, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.3.6.6 to permit the traversing incore 
probe (TIP) system to be considered operable with less than four 
operable TIP units.Date of publication of individual notice in Federal 
Register: July 22, 1994 (59 FR 37516) Expiration date of individual 
notice: Comment Period expires August 8, 1994; Notice period expires 
August 22, 1994
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

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.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
SteamElectric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: September 15, 1993
    Brief description of amendment: The amendment revises the pressure-
temperature limits from 15 to 24 effective full power years.
    Date of issuance: July 29, 1994
    Effective date:  July 29, 1994
    Amendment No. 149
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 13, 1993 (58 FR 
52980) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 29, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Hartsville Memorial Library, 
Home and Fifth Avenues, Hartsville, South Carolina 29550

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
North Carolina

    Date of application for amendment: February 4, 1994
    Brief description of amendment: The amendment revises the Action 
Statement of TS 3.6.5, Vacuum Relief System, to require in Modes 1-4 
with one vacuum relief system inoperable that the system be restored to 
the operable status within seventy-two hours or be in at least hot 
standby within the next six hours.
    Date of issuance: July 27, 1994
    Effective date: July 27, 1994
    Amendment No. 49
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14886) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 27, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
North Carolina

    Date of application for amendment: May 11, 1994
    Brief description of amendment: The amendment revises TS 3/4.2.3 to 
establish limits on reactor power level as a function of total reactor 
coolant system (RCS) flow rate up to 5 percent below the current 
specified flow rate.
    Date of issuance: July 27, 1994
    Effective date: July 27, 1994
    Amendment No. 50
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27079) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 27, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

    Date of application for amendments: March 30, 1994, as supplemented 
by letters dated June 13, June 14, July 11, July 21 and July 28, 1994
    Brief description of amendments: The amendment revises the 
Technical Specifications (TSs) by changing the Unit 1 heatup and 
cooldown pressure-temperature (P-T) curves (i.e., Figures 3.4-2a and 
3.4-3a) to incorporate a newly determined reactor pressure vessel (RPV) 
reference nil-ductility temperature, RTNDT. This new value of 
RTNDT was determined from the licensee's analysis of the first 
irradiation sample removed from Unit 1. The setpoint curve contained in 
Figure 3.4-4a for the Unit 1 Low Temperature Overpressure Protection 
System (LTOPS) is also revised to reflect the changes in the P-T curves 
and to provide a margin for uncertainties in measuring the reactor 
pressure. Additionally, the amendment updates the removal schedule of 
RPV surveillance capsules for both units in accordance with the 
American Society for Testing Materials (ASTM) Standard ASTM E185-82. 
Finally, the amendment incorporates an editorial change for Unit 2 in 
which some clarifying text was added in the Table of Contents to 
indicate the lifetime applicability of Figure 3.4-4b for Unit 2.
    Date of issuance: July 29, 1994
    Effective date: July 29, 1994
    Amendment Nos.: 53 and 53
    Facility Operating License Nos. NPF-72 and NPF-77. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1994 (59 FR 
24747) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 29, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Wilmington Township Public 
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: September 10, 1993 as 
supplemented November 17, 1993
    Brief description of amendments: The amendments revise the LaSalle 
County Station, Units 1 and 2 Updated Final Safety Analysis Report 
Section 11.5.2.1.4 to specify that operator action is required to trip 
the mechanical vacuum pump upon receipt of a main steam line high 
radiation alarm, rather than the action of an automatic trip, which is 
currently described in the UFSAR. NRC approval was required because the 
required operator action, an existing condition, is contrary to that 
described in the UFSAR and the NRC's Safety Evaluation Report related 
to the operation of LaSalle County station (NUREG-0519), and involved 
an unreviewed safety question.
    Date of issuance: July 26, 1994
    Effective date: July 26, 1994
    Amendment Nos.: 101 and 85
    Facility Operating License Nos. NPF-11 and NPF-18. The amendments 
revised the UFSAR.
    Date of initial notice in Federal Register: December 1, 1993 (58 FR 
63403) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 26, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Public Library of Illinois 
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, Illinois

    Date of application for amendments: June 16, 1994
    Brief description of amendments: The amendments change 
specification 3/4.10.1 to recognize the exemption of a single valve on 
each unit from Type C testing until the next refueling outage on each 
unit.
    Date of issuance: August 1, 1994
    Effective date: August 1, 1994
    Amendment Nos.: 155 and 143
    Facility Operating License Nos. DPR-39 and DPR-48. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1994 (59 FR 
33798) Public comments requested as to proposed no significant hazards 
consideration: yes. The 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 an opportunity to request a hearing by August 1, 1994, 
but indicated that if the Commission makes a final no significant 
hazards consideration determination, any such hearing would take place 
after issuance of the amendment. The Commission's related evaluation of 
the amendment and final significant hazards consideration determination 
is contained in a Safety Evaluation dated August 1, 1994.
    Local Public Document Room location:  Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

Consolidated Edison Company of New York, Docket No. 50-003 and 
Docket No. 50-247, Indian Point Nuclear Generating Unit Nos. 1 and 
2, Westchester County, New York

    Date of application for amendments: September 29, 1993
    Brief description of amendments: The amendments revise the 
Technical Specifications (TSs) to change the submittal frequency of the 
Radioactive Effluent Release Report from semiannually to annually, and 
change the reporting date.
    Date of issuance: July 21, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 44 and 172
    Facility Operating License Nos. DPR-5 and DPR-26: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62153) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 21, 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.

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: September 29, 1993, as 
supplemented by letter dated April 1, 1994.
    Brief description of amendment: The amendment revises the Technical 
Specifications to remove the cycle-specific parameter limits and to 
reference a Core Operating Limits Report containing these limits. These 
changes are in accordance with the guidance provided in Generic Letter 
88-16, ``Removal of Cycle-Specific Parameter Limits from Technical 
Specifications.''
    Date of issuance: July 26, 1994
    Effective date: This license amendment is effective as of the date 
of issuance of the COLR by the licensee to be implemented no later than 
the return to operation following the 1995 refueling outage.
    Amendment No.: 173
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62154) The April 1, 1994, provided additional information that did 
not change the initial determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
July 26, 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.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: January 28, 1994
    Brief description of amendment: The amendment revises the TSs to 
change the containment isolation valve testing frequency and the 
acceptance criteria for the combined containment leakage rate to 
accommodate operation on a 24-month fuel cycle. These changes follow 
the guidance provided in Generic Letter 91-04, ``Changes in Technical 
Specification Surveillance Intervals to Accommodate a 24-Month Fuel 
Cycle.''
    Date of issuance: July 29, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 174
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17596) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 29, 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: April 22, 1994, as supplemented 
July 6, 1994
    Brief description of amendment: The amendment revised the reactor 
vessel pressure-temperature limits in the Technical Specifications. The 
change insures that the vessel fracture toughness requirements of 
Section V of 10 CFR Part 50, Appendix G, are satisfied through end of 
life.
    Date of issuance: July 25, 1994
    Effective date: July 25, 1994
    Amendment No.: 113
    Facility Operating License No. DPR-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1994 (59 FR 
24749). The July 6, 1994, letter provided clarifying information within 
the scope of the initial notice and did not affect the staff's proposed 
no significant hazards consideration findings. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
July 25, 1994.No significant hazards consideration comments received: 
No.
    Local Public Document Room location: North Central Michigan 
College, 1515 Howard Street, Petoskey, Michigan 49770.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, 
Michigan

    Date of application for amendment: May 10, 1994
    Brief description of amendment: The amendment revises the Fermi-2 
Technical Specifications (TS) to remove Table 3.6.3-1, the list of 
primary containment isolation valves and Table 3.8.4.3-1, the list of 
safety systems' motor-operated valve thermal overload protection from 
the TS to administrative procedures in accordance with the guidance 
contained in Generic Letter 91-08.
    Date of issuance: August 1, 1994
    Effective date: August 1, 1994, with full implementation within 45 
days.
    Amendment No.: 102
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29626) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 1, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.

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: November 11, 1993, as 
supplemented on June 13, 1994
    Brief description of amendments: The amendments revise the 
Technical Specification surveillance requirements for the emergency 
core cooling system subsystems.
    Date of issuance: July 29, 1994
    Effective date: July 29, 1994
    Amendment Nos.: 145 and 127
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17579) The June 13, 1994, letter provided clarifying and additional 
information that did not change the scope of the November 11, 1993, 
application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated July 29, 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

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 5, 1994, as supplemented 
June 13, 1994.
    Brief description of amendments: The amendments revise the 
Technical Specifications to increase Main Steam and Pressurizer Code 
Safety Valve Setpoint Tolerances.
    Date of issuance: August 2, 1994
    Effective date: August 2, 1994
    Amendment Nos.: 146 and 128

    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1994 (59 FR 
32029) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 2, 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

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam 
ElectricStation, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: December 14, 1993
    Brief description of amendment: The amendment revised the Technical 
Specifications to revise the azimuthal power tilt limit from less than 
or equal to 0.10 (10%) to less than or equal to 0.03 (3%) and revises 
the action statement for control element assembly misalignment to allow 
24 hours to restore the tilt to less than 3%.
    Date of issuance: August 3, 1994
    Effective date: August 3, 1994
    Amendment No.: 97
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 19, 1994 (59 FR 
2866) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 3, 1994.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: April 6, 1994, as supplemented 
June 21, 1994
    Brief description of amendment: The amendment eliminates the scram 
and main steam line isolation valve (MSIV) closure requirements 
associated with the main steam line radiation monitors (MSLRM). The 
amendment also eliminates the following related automatic isolation 
functions that are associated with the MSLRM scram and MSIV isolation: 
a) Main Steam Line Condenser Drain Valves, b) Emergency Condenser Drain 
Valves, c) Reactor Recirculation Loop Sample Valve, d) Instrumental Air 
Valves, and e) Condenser Pump Isolation.
    Date of issuance: July 29, 1994
    Effective date: As of the date of issuance to be implemented at the 
restart from refueling outage 15R.
    Amendment No.: 169
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1994 (59 FR 
22008). The June 21, 1994, letter provided 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 July 29, 
1994.No significant hazards consideration comments received: No.
    Local Public Document Room location:  Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, New Jersey 
08753.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile 
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of application for amendment: July 15, 1993.
    Brief description of amendment: The amendment revises the plant 
Technical Specifications (TSs) on the Reactor Coolant Inventory 
Trending System (RCITS). The change is consistent with NUREG-1430 
entitled ``Standard Technical Specifications for Babcock and Wilcox 
Plants.'' The RCITS information will be available to the operator to 
enhance the operator's ability to understand and manage transients and 
events when needed.
    Date of issuance: August 1, 1994
    Effective date: As of its date of issuance, to be implemented 
within 30 days of issuance.
    Amendment No.: 191
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications. I11Date of initial notice in Federal 
Register: June 8, 1994 (59 FR 29626) The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
August 1, 1994. No significant hazards consideration comments received: 
No.
    Local Public Document Room location:  Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile 
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of application for amendment: February 10, 1994
    Brief description of amendment: The amendment revises the TMI-1 
Technical Specifications (TS) to revise specification 3.7.2.c, ``Unit 
Electric Power System,'' to provide an option to testing an emergency 
diesel generator (EDG) when the redundant EDG is inoperable.
    Date of issuance: July 25, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 188
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 22, 1994 (59 FR 
32230) The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated July 25, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile 
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of application for amendment: March 2, 1994
    Brief description of amendment: The amendment revises the plant 
Technical Specifications to modify Operational Safety Instrumentation 
requirements to specify completion time which allows for performance of 
maintenance or surveillance within a reasonable time and to be 
consistent with the allowable outage time for other safety-related 
equipment when only one train is affected.
    Date of issuance: July 25, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days after issuance.
    Amendment No.: 189
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17600). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 25, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile 
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of application for amendment: March 11, 1994
    Brief description of amendment: The amendment revises the plant 
Technical Specifications to specify an allowable outage time for the 
Emergency Feedwater Pumps during surveillance activities. It also 
changes the requirement to test redundant components for operability to 
a requirement to ensure operability based on verification of completion 
of appropriate surveillance activities.
    Date of issuance: July 25, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days of issuance.
    Amendment No.: 190
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17601).The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 25, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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: January 14, 1994
    Brief description of amendment: The amendment revised TS Sections 
3/4.3, ``Instrumentation,'' 3/4.4.2, ``Safety/Relief Valves,'' and 
associated Bases to increase the surveillance test intervals and 
allowable out-of-service times for various instruments.
    Date of issuance: August 2, 1994
    Effective date: August 2, 1994
    Amendment No.: 74
    Facility Operating License No. NPF-47. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1994 (59 FR 
21787) The additional information contained in the supplemental letter 
dated July 15, 1994, was clarifying in nature and thus, within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination.The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated August 2, 1994.No significant hazards consideration comments 
received. No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, 
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, 
Matagorda County, Texas

    Date of amendment request: March 16, 1994
    Brief description of amendments: The amendments modified Figure 
3.4-4, ``Nominal Maximum Allowable PORV Setpoint for the Cold 
Overpressure System,'' for the cold overpressure mitigation system with 
a revised setpoint curve.
    Date of issuance:  August 3, 1994
    Effective date: August 3, 1994, to be implemented within 31 days of 
issuance
    Amendment Nos.:  Unit 1 - Amendment No. 63; Unit 2 - Amendment No. 
52
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17601) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 3, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488

Northeast Nuclear Energy Company, Docket No. 50-245, 
MillstoneNuclear Power Station, Unit 1, New London County, 
Connecticut

    Date of application for amendment: April 29, 1994
    Brief description of amendment: The amendment changes the 
requirement for reactor operators in Table 6.2-1 from 2 to 3 for the 
RUN, STARTUP/HOT STANDBY and HOT SHUTDOWN conditions. In addition, two 
typographical corrections were made to page 6-4.
    Date of issuance: August 2, 1994
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 75
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 22, 1994 (59 FR 
32231) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 2, 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, Connecticut 06360.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay 
Power Plant, Unit 3, Humboldt County, California

    Date of application for amendment:  October 8, 1993 (Reference HBL-
93-058)
    Brief description of amendment: This amendment modified the 
Technical Specifications (TS) incorporated in
    Facility Operating License No. DPR-7 as Appendix A by incorporating 
a title change into Section VII, Administrative Controls. This change 
reflects a plant organizational name change.
    Date of issuance: July 26, 1994
    Effective date: This license amendment is effective as of the date 
of its issuance and must be fully implemented no later than 30 days 
from the date of issuance.
    Amendment No.: 27
    Facility Operating License No. DPR-7: The amendment revised the TS.
    Date of initial notice in Federal Register: January 5, 1994 (59 FR 
624) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated July 26, 1994No significant hazards 
consideration comments received: No.
    Local Public Document Room location:  Humboldt County Library, 636 
F Street, Eureka, California 95501

Philadelphia Electric Company, Public Service Electric and Gas 
CompanyDelmarva 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: October 27, 1993, as 
supplemented by letters dated April 29, 1994, and June 27, 1994
    Brief description of amendments: These amendments revise the Unit 2 
and Unit 3 Technical Specifications to allow one of the required on-
shift senior reactor operators (SRO) to be combined with the required 
shift technical advisor (STA) position (i.e., dual-role SRO/STA 
position) as long as a minimum of three qualified individuals fill the 
SRO and STA positions.
    Date of issuance: August 2, 1994
    Effective date: August 2, 1994
    Amendments Nos.: 191 and 196
    Facility Operating License Nos. DPR-44 and DPR-56: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64613)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 2, 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: May 10, 1994
    Brief description of amendment: The Technical Specifications 
amendment revised Section 3.1.C.3 and Table 4.1-1 of Appendix A of the 
Operating License. These changes require that the reactor coolant 
average temperature (Tavg) be no lower than 540 deg.F during 
critical operation. Critical operation at Tavg less than 540 deg.F 
requires operator response to restore Tavg to greater than or 
equal to 540 deg.F within 15 minutes or be in hot shutdown within the 
following 15 minutes. Additionally, the change in Table 4.1-1 entitled, 
``Minimum Frequencies for Checks, Calibrations and Tests,'' adds the 
requirement for Tavg instrument check frequency to be reduced to 
30 minutes when the Tavg banks are above zero steps. Furthermore, 
the revision to the Bases indicates that the minimum temperature for 
criticality provides assurance that the reactor is operated within the 
bounds of the safety analyses. Also included is an administrative 
change to correct some typographical errors on page 3.1-25 of the 
Technical Specifications.
    Date of issuance: July 25, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 149
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29630) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 25, 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.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
County, New Jersey

    Date of application for amendments: May 3, 1994
    Brief description of amendments: The amendments revised the 
Technical Specification (TS) for Combustible Gas Control (3/4.6.4.1) by 
changing the surveillance frequency for performing the channel 
functional test to once-per-quarter and the channel calibration to 
once-per-refueling. Also, the TS for the Auxiliary Feedwater System (3/
4.7.1.2) were changed to reduce the surveillance frequency for 
performing pump operability tests to once-per-quarter on a staggered 
test basis. These changes are consistent with the provisions of Generic 
Letter 93-05, ``Line-Item Technical Specifications Improvements to 
Reduce Surveillance Requirements For Testing During Power Operations.''
    Date of issuance: July 27, 1994
    Effective date: July 27, 1994
    Amendment Nos. 153 and 134
    Facility Operating License Nos. DPR-70 and DPR-75. These amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29634) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 27, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

South Carolina Electric & Gas Company, South Carolina Public 
ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear 
Station, Unit No. 1, Fairfield County, South Carolina

    Date of application for amendment: March 11, 1994
    Brief description of amendment: The amendment changes the Technical 
Specifications to delete TS Surveillance Requirement 4.8.4.1.a.3 that 
requires periodic retest of containment penetration overcurrent 
protection fuses and to remove references to containment penetration 
fuse testing from the TS Bases.
    Date of issuance: July 29, 1994
    Effective date: July 29, 1994
    Amendment No.: 115
    Facility Operating License No. NPF-12. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1994 (59 FR 
24752) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 29, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Fairfield County Library, 
Garden and Washington Streets, Winnsboro, South Carolina 29180.

Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph 
M. Farley Nuclear Plant, Unit 1, Houston County, Alabama.

    Date of application for amendment: June 17, 1994
    Brief description of amendment: The amendment changes the Technical 
Specifications to revise the nuclear enthalpy rise hot channel factor 
(F delta H) from equal to or less than 1.65 [1 plus 0.3(1-P)] to equal 
to or less than l.70 [1 plus 0.3(1-P)] where P is a fraction of rated 
power. The amendment also revises the action statement to reflect 
guidance contained in the improved standard technical specifications.
    Date of issuance: July 22, 1994
    Effective date: July 22, 1994
    Amendment No.: 109
    Facility Operating License No. NPF-2. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 22, 1994 (59 FR 
32249) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 22, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, 
Alabama

    Date of application for amendments: August 25, 1992 (TS 321)
    Brief description of amendments: The amendments delete reference to 
recirculation equalizer valves from the technical specifications. These 
components have been removed from Browns Ferry Unit 3, and are not used 
in Browns Ferry Units 1 and 2.
    Date of issuance: August 4, 1994
    Effective date: August 4, 1994
    Amendment Nos.: 211, 226 and 184
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 28, 1992 (57 FR 
48829) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 4, 1994.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: June 17, 1993 (TS 93-08)
    Brief description of amendments: The amendments revise the 
allowable values for the intermediate and source range neutron flux 
reactor trip setpoints.
    Date of issuance: July 26, 1994
    Effective date: July 26, 1994
    Amendment Nos.: 185 and 177
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41514) The Commission's related evaluation of the amendments are 
contained in a Safety Evaluation dated July 26, 1994.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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, Ohio

    Date of application for amendment: February 17, 1993
    Brief description of amendment: This amendment increases the TS 
trip setpoint and its associated allowable value for containment high-
radiation specified in TS Table 3.3-4 from ``<2 x Background at RATED 
THERMAL POWER'' to ``<4 x Background at RATED THERMAL POWER.''
    Date of issuance: July 27, 1994
    Effective date: As of the date of issuance to be implemented within 
90 days.
    Amendment No. 190
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 23, 1993 (58 FR 
34096) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 27, 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: February 14, 1994, as supplemented by 
letter dated April 29, 1994.
    Brief description of amendments: The amendments increase the boron 
concentration limits for the Unit 2 refueling water storage tank and 
emergency core cooling system, and delete a footnote concerning 
refueling canal boron concentration during initial fuel load for both 
units.
    Date of issuance: August 2, 1994
    Effective date: August 2, 1994, to be implemented prior to startup 
for Cycle 2 for Comanche Peak Steam Electric Station, Unit 2.
    Amendment Nos.: Unit 1 - Amendment No. 26; Unit 2 - Amendment No. 
12
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1994 (59 FR 
22015) The information contained in the April 29, 1994, letter was 
editorial in nature and thus, within the scope of the initial notice 
and did not affect the staff's proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated August 2, 1994.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O. 
Box 19497, Arlington, Texas 76019.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: February 17, 1994, supplemented 
by letter dated May 18, 1994
    Brief description of amendment: The amendment revises the Technical 
Specification 3/4.5.1 and associated Bases Section 3/4.5.1. A new 
Action Statement a. provides a 72-hour allowed outage time (AOT) for 
one accummulator inoperable due to boron concentration. The Action 
Statement b. AOT was changed to 24 hours. Surveillance Requirements 
4.5.1.1.a.1 and 4.5.1.1.b were revised and 4.5.1.2 was deleted from the 
TS.
    Date of issuance: August 5, 1994
    Effective date: August 5, 1994 to be implemented within 30 days
    Amendment No.: 91
    Facility Operating License No. NPF-30. Amendment revised the 
Technical Specifications 3/4.5.1 and associated Bases Section 3/4.5.1.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14898) The additional information contained in the May 18, 1994, letter 
provided additional supplemental information that did not change the 
initial proposed no significant hazards consideration. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated August 5, 1994.No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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 19, 1994
    Brief description of amendments: The amendments revise the NA-1&2 
Technical Specifications surveillance frequency requirements for 
control rod motion testing from once per 31 days to once per 92 days.
    Date of issuance: July 28, 1994
    Effective date: July 28, 1994
    Amendment Nos.: 185 and 166
    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 
27070) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 28, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

    Date of application for amendments: April 19, 1994
    Brief description of amendments: These amendments modify the 
surveillance frequency of the control rod motion testing from monthly 
to quarterly
    Date of issuance: August 2, 1994
    Effective date: August 2, 1994
    Amendment Nos. 192 and 192
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27070) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 2, 1994No significant 
hazards consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of application for amendment: April 1, 1993
    Brief description of amendment: The amendment modifies the 
Technical Specifications to add inservice inspection requirements for 
reactor coolant system piping in accordance with Generic Letter 88-01, 
``NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in 
BWR Austenitic Stainless Steel Piping.'' In addition, the amendment 
corrects an administrative error in a TS that references a table 
listing high/low pressure interface valve leakage pressure monitors.
    Date of issuance: July 28, 1994
    Effective date: 30 days after the date of issuance.
    Amendment No.: 130
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1993 (58 FR 
28065) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated July 28, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of application for amendment: February 17, 1994, supplemented 
by letter dated May 13, 1994.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TS) 10-year hydrostatic testing requirements. The 
changes (1) add a special test exception for inservice leak testing and 
hydrostatic testing, (2) add a new minimum reactor vessel metal 
pressure-temperature curve for less than or equal to eight effective 
full power years, and (3) delete Table B 3/4.4.6-1, ``Reactor Vessel 
Toughness,'' from the TS bases.
    Date of issuance: May 27, 1994
    Effective date: May 27, 1994
    Amendment No.: 122
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14902) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 27, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location:  Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
NuclearPower Plant, Kewaunee County, Wisconsin

    Date of application for amendment: December 1, 1993
    Brief description of amendment: The amendment revises the Kewaunee 
Nuclear Power Plant (KNPP) Technical Specifications (TS) by 
incorporating technical and administrative changes to TS 3.10, Control 
Rod and Power Distribution Limits. The changes eliminate specifications 
for fuel designs no longer used at Kewaunee, specify required actions 
to be taken upon exceeding control bank insertion limits, and revise 
the limits for Departure from Nucleate Boiling (DNB) related parameters 
to assure operation within the assumptions of the Updated Safety 
Analysis Report (USAR) analyses.
    Date of issuance: August 3, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 110
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4949) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 3, 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.

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 September 16, 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. 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).

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Date of application for amendments: August 2, 1994
    Brief description of amendments: The amendments revise the 
Technical Specifications by adding a footnote that recognizes that 
through the end of cycle 6, the Unit 1, loop B wide range hot leg 
indication at the remote shutdown panel is inoperable.
    Date of issuance: August 5, 1994
    Effective date: August 5, 1994
    Amendment Nos.: 63 and 63
    Facility Operating License Nos. NPF-37 and NPF-66. The amendments 
revised the Technical Specifications.Public comments requested as to 
proposed no siginificant hazards consideration: No. The Commission's 
related evaluation of the amendments, finding of emergency 
circumstances, and final determination of no significant hazards 
consideration are contained in a Safety Evaluation dated August 5, 
1994.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    Local Public Document Room location: Byron Public Library, 109 N. 
Franklin, P.O. Box 434, Byron, Illinois 61010.
    NRC Project Director: Robert A. Capra
    Dated at Rockville, Maryland, this 10th day of August 1994.
    For The Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II Office of Nuclear Reactor 
Regulation
[Doc. 94-20006 Filed 8-16-94 8:45 am]
BILLING CODE 7590-01-F