[Federal Register Volume 59, Number 41 (Wednesday, March 2, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-4562]


[[Page Unknown]]

[Federal Register: March 2, 1994]


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NUCLEAR REGULATORY COMMISSION
Biweekly Notice

 

Applications and Amendments to Facility Operating Licenses 
Involving No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from February 5, 1994, through February 17, 1994. 
The last biweekly notice was published on February 16, 1994 (59 FR 
7685).

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 P-223, Phillips Building, 7920 Norfolk Avenue, 
Bethesda, 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 April 1, 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: January 13, 1994
    Description of amendment requests: Request for NRC consent to the 
indirect transfer of control of El Paso Electric Company's interest in 
Operating License Nos. NPF-41, NPF-51, NPF-74 and to amend Operating 
License Nos. NPF-51 and NPF-74 to delete provisions for El Paso 
Electric Company's sale-leaseback arrangements.
    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.
    This amendment request does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because the proposed change is administrative in nature. 
The proposed change deletes Sections 2.B.(7)(a) and (b) of License 
No. NPF-51, and Sections 2.B.(6)(a) and (b) of License No. No. NPF-
74. These section describe the structure of the financing of El 
Paso's interest in Palo Verde, specifically authorizing sale and 
leaseback transactions. The proposed change does not affect the 
assumptions used in the accident analyses, nor does the proposed 
change result in changes to the physical configuration of the 
facility, design parameters, technical specifications, or operation 
and maintenance of the facility. Therefore, this amendment request 
does 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 analyzed.
    This amendment request does not create the possibility of a new 
or different kind of accident from any accident previously analyzed 
because the proposed change is administrative in nature. The 
proposed change deletes Sections 2.B.(7)(a) and (b) of License No. 
NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74. These 
sections describe the structure of the financing of El Paso's 
interest in Palo Verde, specifically authorizing sale and leaseback 
transactions. The proposed change does not involve modifications to 
any of the existing equipment nor does the change affect the 
operation and maintenance of the facility. Therefore, this amendment 
request does not create the possibility of a new or different kind 
of accident not previously analyzed.
    Standard 3 -- Involve a significant reduction in a margin of 
safety.
    This amendment request does not involve a significant reduction 
in a margin of safety because it is administrative in nature. The 
proposed change deletes Sections 2.B.(7)(a) and (b) of License No. 
NPF-51, and Sections 2.B.(6)(a) and (b) of License No. NPF-74. These 
sections describe the structure of the financing of El Paso's 
interest in Palo Verde, specifically authorizing sale and leaseback 
transactions. The proposed change does not involve changes to any 
existing plant equipment or accident analyses that provide for or 
establish margins of safety. There are no changes to the operation 
or maintenance of the facility and the existing margins of safety 
are not changed by the proposed change. Therefore, this amendment 
request does not involve a signigicant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensees' analysis and, based on 
that review, it appears that the proposed license amendment reflects 
only a change in the structure of the financing of El Paso's interest 
in Palo Verde and 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

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

    Date of amendment request: February 4, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3/4.6.4, Containment Systems Combustible 
Gas Control, by eliminating the 12-hour channel check surveillance 
requirement for the containment hydrogen monitoring system in 
conformance with the new Standard Technical Specifications, NUREG-1431.
    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.
    Final Safety Analysis Report [FSAR] section 6.2.5.2.3 states 
that the Hydrogen Analyzer is only required to be functioning 
(continuously indicating and recording hydrogen concentration) 
within 30 minutes of safety injection initiation. The performance of 
an analog operational test every 31 days and a channel calibration 
test every 92 days verifies this operability. Based on this, the 
monitors will be fully capable of performing their intended design 
function following a safety injection initiation. Therefore, the 
elimination of the 12-hour channel check would not increase the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The Hydrogen Monitors perform an ``indication'' function only, 
[sic] to help ensure that hydrogen concentrations within containment 
are maintained below flammable limits during a post-LOCA [loss-of-
coolant accident] condition. The proposed changes do not involve any 
modifications or additions to plant equipment and the design and 
operation of the plant will not be affected. Therefore, the 
elimination of the 12-hour channel check does not affect any 
parameters which relate to the margin of safety as defined in the 
Technical Specifications or the FSAR. Therefore, the proposed 
changes do not involve a significant reduction in a margin of 
safety.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The proposed elimination of the 12-hour channel check does not 
affect any parameters which relate to the margin of safety as 
defined in the Technical Specifications or the FSAR. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: S. Singh Bajwa

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of amendment request: March 26, 1993
    Description of amendment request: The proposed amendment would 
modify trip level settings for the Isolation Condenser and High 
Pressure Core Injection (HPCI) System Steam lines to more conservative 
values. In addition, the proposed amendment would revise the Emergency 
Core Cooling System Low-Low Water Level initiation trip level setting 
tolerance.
    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:
    Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    HPCI Steamline High Flow Isolation Trip Level Setting
    The purpose of the HPCI leak detection systems are to detect 
breaks in the system piping. Normal steam flows within the system 
can fluctuate in excess of 250% rated flow and exceed 500% rated 
steam flow after experiencing a break. During the original licensing 
of the plant, it was analytically determined by GE that three times 
maximum steam flow (300%) is the optimum setpoint for the isolation 
of HPCI. A 300% steam flow setpoint ensures that spurious trips are 
avoided and that breaks in the piping are identified. Because the 
HPCI High Steamline Flow Isolation setpoint is not assumed as an 
accident precursor, the probability of any previously evaluated 
accident is not increased by the changed setpoint.
    The proposed changes to the setpoint allow a more accurate and 
conservative value for 300% steam flow. The proposed change in 
conjunction with a more conservative field setting ensures HPCI 
isolation occurring between the range of 300% and 500% steam flow, 
thus ensuring HPCI isolation in the event of a pipe break. Because 
the HPCI high steamline flow setpoint will be maintained above 
normally found operational values (272% steam flow) and below 
expected conditions with a pipe break (500% steam flow), the 
consequences of any previously evaluated accident are not increased 
with the proposed setpoint change.
    solation Condenser Steamline High Flow Isolation Trip Level 
Setting
    The purpose of the Isolation Condenser leak detection 
instrumentation is to detect breaks in the system piping. Normal 
steam flows within the system can fluctuate in excess of 250% rated 
flow and exceed 500% rated steam flow after experiencing a break. 
During the original licensing of the plant, it was analytically 
determined by GE that three times rated steam flow (300%) is the 
optimum setpoint for the isolation of the Isolation Condenser. A 
300% steam flow Isolation setpoint ensures that spurious trips are 
avoided and that breaks in the piping are identified. Because the 
Isolation Condenser High Steamline Flow setpoint is not assumed as 
an accident precursor, the probability of any previously evaluated 
accident is not increased by the changed setpoint. The proposed 
changes to the setpoint provide a more accurate and conservative 
field setting for 300% steam flow.
    The proposed changes in conjunction with a more conservative 
field setting results in Isolation Condenser isolation occurring 
between the range of 300% and 500% steam flow, thus ensuring 
Isolation Condenser isolation in the event of a pipe break. Because 
the Isolation Condenser High Steamline Flow Isolation setpoint will 
be maintained above normally found operational values (272% steam 
flow) and below expected conditions with a pipe break (500% steam 
flow), the consequences of any previously evaluated accident are not 
increased with the proposed setpoint change.
    Reactor Low-Low Water Level Trip Level Setting Tolerance
    The Low-Low Reactor Water Level trip setting is designed to 
initiate ECCS when reactor water level is less than or equal to 444 
inches above vessel zero. Top of active fuel (TAF) is defined as 360 
inches above vessel zero. -59 inches is 84 inches above TAF. The 
present trip setting tolerance (84 inches, + 4, - 0, above TAF) only 
allows a deviation of 4 inches in the conservative direction. The 
proposed change (greater than or equal to 84 inches) does not impose 
a restriction on the limit toward the conservative direction. 
Because a level switch trip level setting by itself is not assumed 
as an accident precursor, the probability of any previously 
evaluated accident is not increased by the changed setpoint.
    The proposed change eliminates a restriction on the trip level 
setting for Low-Low Reactor Water Level. Dresden proposes modifying 
the acceptance limit of the Low-Low trip setting such that the 
instrument field setting will not deviate below 84 inches. 
Therefore, the actuation of appropriate ECCS are unchanged and the 
consequences of any previously evaluated accident are not increased 
with the proposed setpoint change.
    Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    HPCI Steamline High Flow Isolation Trip Level Setting
    The purpose of the HPCI Steamline High Flow Isolation trip level 
setting is to detect breaks in system piping and initiate isolation 
of the system if breaks are discovered. Normal steam flows within 
the system can fluctuate as high as 250% rated flow and exceed 500% 
rated steam flow after experiencing a break. 300% steam flow has 
been used as the setpoint to ensure that spurious trips are avoided 
and that breaks in the piping are identified. The changes to the 
HPCI High Steamline Flow setpoint ensure that isolation occurs at 
300% rated steam flow (below 500% rated steam flow). The current 
setpoint will also isolate below 500% rated steam flow. Because the 
new setpoint continues to allow normal operational flexibility and 
ensures isolation in the event of a pipe break, the proposed changes 
do not create the possibility of a new or different kind of accident 
than previously evaluated.
    Isolation Condenser Steamline High Flow Isolation Trip Level 
Setting
    The purpose of the Isolation Condenser Steamline High Flow 
Isolation trip level setting is to detect breaks in system piping 
and initiate isolation of the system if breaks are discovered. 
Normal steam flows within the system can fluctuate in excess of 250% 
rated flow and exceed 500% rated steam flow after experiencing a 
break. 300% steam flow has been used as the setpoint to ensure that 
spurious trips are avoided and ensures that isolation occurs at 300% 
rated steam flow (below 500% rated steam flow). The current setpoint 
will also isolate below 500% rated steam flow. Because the new 
setpoint continues to allow normal operational flexibility and 
ensures isolation in the event of a pipe break, the proposed changes 
do not create the possibility of a new or different kind of accident 
than previously evaluated.
    Reactor Low-Low Level Trip Level Setting Tolerance
    The Reactor Low-Low Water Level trip setting is designed to 
initiate the appropriate ECCS when Reactor Water Level is 
decreasing. The proposed change to the setpoint only eliminates the 
overly burdensome restriction within the setpoint tolerances. The 
absolute low limit of 84 inches is unchanged, thus maintaining all 
assumptions related to 84 inches (-59 inches indicated level) within 
Dresden's Safety Analysis. The removal of the upper tolerance will 
not increase the probability of inadvertent ECCS initiation since 
the actual field setting will be at a reactor vessel level which has 
not been reached in 40 + years of operation at Dresden Units 2 and 
3. Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident than previously evaluated.
    Involve a significant reduction in the margin of safety because:
    High Pressure Coolant Injection Setpoint
    The HPCI high steamline flow setpoint ensures that isolation 
occurs at 300% maximum steam flow (below 500% rated steam flow). The 
current Technical Specification setpoint will also allow isolation 
below 500% rated steam flow but at a value greater than 300%. Thus, 
the proposed setpoint isolates at a lower steam flow rate than the 
current limit. Therefore, because isolation of HPCI would occur at a 
lower steam flow rate during a pipe break, the proposed changes do 
not involve a significant reduction in the margin of safety.
    Isolation Condenser Steamline High Flow Isolation Trip Level 
Setting
    The Isolation Condenser High Steamline Flow Isolation Trip Level 
setting ensures that isolation occurs at 300% rated steam flow 
(below 500% rated steam flow). The current setpoint will also 
isolate below 500% rated steam flow but at a value greater than 
300%. Thus, the proposed setpoint isolates at a lower steam flow 
rate than the current limit. Therefore, because isolation of the 
Isolation Condenser would occur at a lower steam flow rate during a 
pipe break, the proposed changes do not involve a significant 
reduction in the margin of safety.
    Reactor Low-Low Level Trip Level Setting Tolerance
    The Reactor Low-Low Water Level trip setting tolerance ensures 
the proper initiation of appropriate ECCS in the event of a loss of 
inventory to the vessel. The proposed change to the setpoint only 
eliminates the restriction within the setpoint tolerances. The 
absolute low limit of 84 inches is unchanged, thus maintaining all 
assumptions related to 84 inches (minus 59 inches indicated) within 
Dresden's Safety Analysis. Therefore, the proposed changes do not 
involve a significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Morris Public Library, 604 
Liberty Street, Morris, Illinois 60450
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: James E. Dyer

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: December 20, 1993
    Description of amendment request: The proposed amendment would 
revise a minimum critical power ratio (MCPR) safety limit from 1.06 to 
1.07 based on General Electric Standard Application for Reactor Fuel II 
(GESTAR II) NEDE-24011-P-A-10 for GE10 fuel design. The NRC staff has 
previously reviewed and approved the GE10 fuel design.
    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 increase in 
the probability or consequences of an accident previously evaluated 
because:
    The change is based on GE's generic rule licensing document 
GESTAR II (NEDE-24011-P-A-10) which has conservatively addressed the 
use of GE10 fuel in D-lattice cores with NRC approved methods and 
therefore does not adversely affect the consequences of previously 
evaluated accidents. The Safety Limit MCPR change does not affect 
the probability of analyzed accidents because it does not adversely 
impact any equipment important to safety. Increasing the Safety 
Limit MCPR from 1.06 to 1.07 upon implementation of GE10 fuel for 
Cycle 14 operation of Quad Cities Units 1 and 2 therefore does not 
involve a significant increase in the probability or consequences of 
any accident previously evaluated in the FSAR.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because:
    The Safety Limit MCPR change results from the use of NRC 
approved methods in GESTAR II NEDE-24011-P-A-10 for application to 
GE10 fuel for Cycle 14 for Quad Cities Units 1 and 2. The Safety 
Limit MCPR change does not result in any new interaction with 
equipment related to the safe shutdown of the plant. The change does 
not adversely impact equipment important to safety and, therefore 
does not create the possibility of a new or different kind of 
accident scenario. Therefore, the Safety Limit MCPR change from 1.06 
to 1.07 in no way creates the possibility of a new or different kind 
of accident scenario from any accident previously evaluated.
    The proposed change does not involve a significant reduction in 
a margin of safety because:
    Since the GE10 design in a D-lattice core has a geometry between 
C-lattice and D-lattice designs and the C-lattice design has a 
higher, more restrictive safety limit MCPR that the D-lattice 
design, the use of C-lattice safety limit MCPR for the GE10 design 
is a conservative approach. The GE10 fuel design has been 
generically analyzed with approved methods per GESTAR II NEDE-24011-
P-A-10 and the use of the 1.07 Safety Limit MCPR value has been 
previously approved as conservative for application to GE10 fuel in 
D-lattice plants such as Quad Cities. Therefore, the proposed change 
to increase the Safety Limit MCPR from 1.06 to 1.07 maintains the 
margin to safety relative to the current level.
    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: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: James E. Dyer

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: December 10, 1993
    Description of amendment request: The proposed amendment request 
would revise the Technical Specifications to amend (1) Section 5.3.A 
(Reactor Core) to allow the use of VANTAGE + fuel with ZIRLO cladding 
and fuel with filler rods to allow fuel reconstitution, and (2) the 
Basis to Section 2.1 (Safety Limit: Reactor Core) to allow the use of 
departure from nucleate boiling (DNB) Correlations applicable to 
VANTAGE + fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    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 consequences of an accident 
previously analyzed is increased due to the proposed changes. As 
discussed in [Letter from Thadani to Tritch, ``Acceptance

    for Referencing of Topical Report WCAP-12610, VANTAGE + Fuel 
Assembly Reference Core Report'' (TAC No. 77258) July 1, 1991] the 
fuel containing ZIRLO clad will meet all the same material and 
mechanical design criteria as the Zircaloy clad fuel. The use of 
approved Westinghouse Methodology for fuel assembly reconstitution 
as documented in [Letter from Thadani to Tritch, ``Acceptance for 
Referencing of Topical Report WCAP-13060-P, Westinghouse Fuel 
Assembly Reconstitution Evaluation Methodology'' (TAC No. M821391), 
March 30, 1993] will ensure that all criteria are met. The change to 
the basis of Section 2.1 more accurately describes DNB methodology 
and application.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any previously evaluated?
    Response:

    The changes will not create the possibility of a new or 
different kind of accident. The proposed changes involve approved 
methodology which have been shown to meet design and safety 
criteria. In addition, approved procedures will be used to implement 
the changes.
    Response:
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    The proposed amendment does not involve a significant reduction 
in the margin of safety. The changes involve the use of approved 
methodology which meet design and safety criteria. The change to the 
Section 2.1 basis is descriptive and will more accurately describe 
the DNB methodology used in conjunction with the use of VANTAGE + 
fuel.
    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

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: September 28, 1993
    Description of amendment request: The requested amendments would 
delete the portion of the 18-month surveillance requirement on the 
autoclosure interlock (ACI) contained in TS 4.5.2.d associated with 
verifying that the decay heat removal system suction isolation valves 
automatically close on a reactor coolant system pressure signal. The 
terms decay heat removal (ND) and residual heat removal (RHR) are used 
interchangeably here. Also, an obsolete footnote to TS 4.5.2.e relating 
to the completion of the first Unit 1 refueling outage is proposed to 
be deleted.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The requested amendments reference Westinghouse topical report 
WCAP-11736-A, ``Residual Heat Removal System Autoclosure Interlock 
Removal Report for the Westinghouse Owners Group'', for the general 
justification and safety analysis for removing the ACI feature from 
the Catawba ND suction isolation valves. This WCAP, which 
specifically covers the Catawba Nuclear station, has been deemed an 
acceptable reference by the NRC for use in making plant-specific 
licensing submittals. Additional Catawba-specific information/ 
improvements and analyses, as required by the WCAP and associated 
NRC safety evaluation, have been either completed or committed to, 
thereby ensuring that the WCAP/SE conclusion that removal of the RHR 
ACI produces a net safety benefit remains valid.
    Criterion 1
    The requested amendments will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The deletion of the RHR ACI was analyzed in the WCAP for 
Callaway Nuclear Station in terms of (1) the frequency of an 
interfacing LOCA, (2) the availability of the RHR system, and (3) 
the effect on overpressure transients. Callaway is the WCAP's 
reference plant for Catawba Units 1 and 2, and a Catawba-specific 
Probabilistic Risk Assessment (PRA) review of the WCAP determined 
that removal of the ND ACI at Catawba will not invalidate the basic 
conclusions of the WCAP. Consequently, the following information 
from the Callaway analysis is considered applicable to Catawba Units 
1 and 2.
    With the removal of the ACI and addition of a control room 
alarm, the probabilistic risk analysis predicts a decrease in the 
frequency of interfacing LOCAs from 1.52E-06/year to 1.16E-06/year, 
a decrease of approximately 24%.
    The availability of the RHR system was analyzed in three phases: 
initiation, short term cooling, and long term cooling. The 
probabilistic analysis indicated that deletion of the RHR ACI has no 
impact on the failure probability for RHR initiation. During short 
term cooling (72 hours after initiation), RHR ACI deletion decreased 
the RHR failure probability by 12%, from 1.64E-02 to 1.44E-02. The 
long term cooling RHR failure probability was calculated to decrease 
by 70%, from 3.91E-02 to 1.17E-02.
    Appendix D of the WCAP presents the analysis used to determine 
the effect of removal of the ACI on overpressurization transients. 
The analysis categorizes the types of initiating events, determines 
their frequency of occurrence, and then identifies the consequences 
of these occurrences both with and without the ACI feature. The 
result is a list of overpressure consequence categories with 
associated failure probabilities (reference the WCAP's Appendix D, 
Tables D-14, -15, and -16). For the charging/safety injection event, 
consequence frequencies increased on the order of 1.0E-12/shutdown 
year. This is an insignificant increase, as the overall consequence 
frequency of the charging/safety injection event is 1.25E-01. 
Likewise, for the letdown isolation with RHR system operable case, 
one frequency category was increased on the order of 1.0E-15. Again, 
this is insignificant when compared with the total frequency of 
these events of 1.25E-01. For the letdown isolation with RHR system 
isolated event, the overall consequence frequency was reduced from 
4.45E-01 to 2.22E-01. This occurs because many spurious closures of 
the RHR isolation valves cause the isolation of letdown. Removing 
the RHR ACI reduces the frequency of this event by approximately 
50%. It is concluded that the removal of the RHR ACI circuitry has 
an insignificant impact on the frequency of overpressurization 
events at Callaway (and thus Catawba) Nuclear Station.
    Criterion 2
    The requested amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. The effect of an overpressure transient at cold shutdown 
conditions will not be altered by removal of the ND ACI function. 
With or without the ACI function, the ND system could be subject to 
overpressrue for which the ND relief valves must be relied upon to 
limit pressure to within ND design parameters. While it is true that 
the ACI initiates an automatic closure of the ND suction/isolation 
valves on high NC system pressure, overpressure protection of the ND 
system is provided by the ND system relief valves and not by the 
suction/isolation valves that isolate the ND system from the NC 
system. (Refer to NUREG-0954, ``Safety Evaluation Report related to 
the operation of Catawba Nuclear Station, Units 1 and 2,'' Section 
5.4.4.3.)
    The purpose of the ACI feature is to ensure that there is a 
double barrier between the ND system and the NC system when the 
plant is at normal operating conditions (i.e., heated and 
pressurized) and not in the ND cooling mode. Thus, the ACI feature 
serves to preclude conditions that could lead to a LOCA outside of 
containment due to operator error. The safety function of the ACI is 
not to isolate the ND system from the NC system when the ND system 
is operating in the decay heat removal mode.
    There are several methods to ensure that there is a double 
barrier between the ND system and the NC system when the plant is at 
normal operating conditions. First, plant operating procedures 
instruct the operators to isolate the ND system during plant heatup. 
Second, the alarm that will be installed as part of this change will 
annunciate in the control room given an open or intermediate valve 
position signal in conjunction with a high NC pressure signal. This 
alarm will alert operators that any of the four suction/isolation 
valves is (are) not fully closed and that double isolation has not 
been achieved. In conjunction with this alarm, operators will be 
trained using an annunciator response procedure to ensure that they 
act to restore double isolation or return to a safe shutdown 
condition. Third, the Open Permissive Interlock (OPI), which is not 
being removed, will prevent the opening of the valves whenever NC 
system pressure is greater than 385.5 psig.
    Since relief valves prevent overpressurization of the ND system 
during shutdown conditions and since several methods are in place to 
ensure that the ND system is isolated from the NC system during 
normal plant conditions, removal of the ACI will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Criterion 3
    The requested amendments will not involve a significant 
reduction in a margin of safety. The ND ACI function is not a 
consideration in a margin of safety in the basis for any technical 
specification. Since the probabilistic analysis of the WCAP for 
Callaway (which is applicable to Catawba as discussed above) 
indicates that the availability of the RHR system is increased with 
the removal of the ACI, overall safety will be increased.
    In addition, similar amendments for other Westinghouse plants in 
the past have been determined to not involve significant hazards 
considerations.
    Based upon the preceding analyses, Duke Power Company concludes 
that the requested amendments do not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Loren R. Plisco, Acting

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: January 27, 1994
    Description of amendment request: The requested amendments delete 
the verification that each upper and lower Containment Purge System 
(VP) supply and exhaust valve actuates to its isolation position on a 
High Relative Humidity (70%) isolation test signal and will 
allow elimination of the humidity control function of the VP System 
humidistats.
    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
    This TS [Technical Specification] amendment will not increase 
the probability or consequences of an accident which has been 
previously evaluated. No physical changes will be made to the plant 
that would impact fuel handling inside containment, therefore, there 
is no increase in the probability of an accident. Control wiring 
changes that remove the humidistats from the [Containment Purge] 
System control circuits will be the only physical change.
    The heaters will be maintained providing additional margin over 
analyzed conditions. For the reasons stated above, there will be no 
increase in the consequences of an accident previously evaluated.
    CRITERION 2
    This proposed TS amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. This proposed TS amendment will not cause any physical 
changes to the plant that will impact the handling of fuel inside 
containment or changes to fuel handling procedures. Because the 
plant will operate the same way it does now, this proposed amendment 
does not create the possibility of any new or different accident 
from any previously evaluated.
    CRITERION 3
    This proposed TS change will not cause a significant reduction 
in the margin of safety. The test method use[d] to evaluate the 
carbon after TS changes 90 ([Unit] 1) and 84 
([Unit] 2) does not consider heater availability. However the 
heaters will be tested and maintained per Technical specification 
4.9.4.2.d.2. Therefore, the relative humidity of the air entering 
the carbon adsorber is never expected to reach 95% [relative 
humidity].
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Loren R. Plisco, Acting

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, 
Unit No. 1 (ANO-1), Pope County, Arkansas

    Date of amendment request: January 13, 1994
    Description of amendment request: This amendment revises the 
specifications governing the reactor protection (RPS). It modifies the 
use of the RPS channel bypass as specified by Technical Specification 
(TS) 3.5.1.3 and revises a note with Table 3.5.1-1, to refer to a more 
appropriate action.
    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 RPS and EFIC [emergency feedwater initiation and control] 
system provide accident mitigation features and are not considered 
to be accident initiators. The accident mitigation features of the 
plant are not affected by the proposed amendment. In any 
configuration allowed by the revised specifications, the trip logic 
instituted on the RPS is at least equivalent to the trip logic 
instituted by placing a channel in channel or maintenance bypass. 
The RPS remains single-failure proof with one channel in channel 
bypass, manually tripped, or with an inoperable function unbypassed 
in the untripped state. Therefore, upon receipt of an initiating 
signal, a single failure will not prevent the proper actuation of 
RPS. Should a channel of RPS contain an inoperable function 
unbypassed in the untripped condition which does not affect an EFIC 
channel, any channel of EFIC may be placed in maintenance bypass. 
RPS and EFIC remain single-failure proof in this configuration.
    Administrative controls are established to ensure that all 
inoperable RPS functions are evaluated for continued operation in 
the untripped state. Upon detection of a failed function in any 
channel of RPS, the administratively controlled condition reporting 
process evaluates the failure and its effect on other systems for 
continued operability. The operator is informed of the continuing 
status of inoperable functions through the use of Station Log 
entries and Plant Status board entries. In addition, during 
operation with an inoperable function in the untripped, unbypassed 
condition, the remaining RPS channel key-lock channel bypass 
switches will be ``Hold Carded'' (tagged) to prevent their operation 
without prior management approval consistent with the requirements 
of TS Table 3.5.1-1. Plant management maintains the responsibility 
to approve continued operation with inoperable functions unbypassed 
in the untripped state to ensure that the plant is operated in the 
safest configuration with regard to the extent of the failure, and 
the plant operating conditions. Prior to placing any channel of RPS 
or EFIC in bypass, the operator checks the status of redundant 
systems for operability and TS compliance and takes the proper 
action as required by existing plant conditions, plant operating 
procedures and TS.
    The clarification to TS 3.5.1.3 which directs the operator to 
the appropriate actions if multiple channels become inoperable, or 
in the event of an inoperable channel or inoperable function 
occurring concurrent with one channel in bypass is considered to be 
administrative in nature. The change to Note 6 of Table 3.5.1-1 
results in the correction of misleading information and directs the 
Operator to place the plant in a safe mode depending on the system 
which is affected by a failure, and is also considered to be 
administrative in nature. The Bases changes add additional 
information to clarify the 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 probability or consequences of equipment important to safety 
malfunctioning will not be increased. In any configuration allowed 
by the revised specifications, the trip logic instituted on the RPS 
is at least equivalent to the trip logic instituted by placing a 
channel in channel bypass. The RPS remains single-failure proof with 
one channel in channel bypass, manually tripped, or with an 
inoperable function unbypassed in the untripped state. Therefore, 
upon receipt of an initiating signal, a single failure will not 
prevent the proper actuation of RPS. Should a channel of RPS contain 
an inoperable function unbypassed in the untripped condition which 
does not affect an EFIC channel, any channel of EFIC may be placed 
in maintenance bypass. RPS and EFIC remain single-failure proof in 
this configuration.
    The clarification to TS 3.5.1.3 which directs the operator to 
the appropriate actions if multiple channels become inoperable, or 
in the event of an inoperable channel or inoperable function 
occurring concurrent with one channel in bypass is considered to be 
administrative in nature. The change to Note 6 of Table 3.5.1-1 is 
also considered to be administrative in nature, in that misleading 
information in the specification has been corrected to an 
appropriate requirement. The Bases changes add additional 
information to clarify the specifications.
    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 RPS and EFIC system have the same capabilities to mitigate 
and/or prevent accidents as they had prior to this proposed change. 
Allowing flexibility in the response to a function failure in one 
channel of RPS allows placing the plant in the safest operating 
condition for the existing plant conditions considering the extent 
of the function failure. Operation of an RPS channel with an 
inoperable function unbypassed in the untripped state results in 
placing the inoperable function in a 2-out-of-3 trip logic 
(equivalent to channel bypass) while the remainder of the RPS 
functions remain in the normal 2-out-of-4 trip logic. The ANO-1 RPS 
has been reviewed as a 3 channel system with one channel in bypass. 
Implementing this change results in additional conservatism with 
respect to any postulated single-failures.
    Administrative controls are established to ensure that all 
inoperable RPS functions are evaluated for continued operation in 
the untripped state. Upon detection of a failed function in any 
channel of RPS, the administratively controlled condition reporting 
process evaluates the failure and its effect on other systems for 
continued operability. The operator is informed of the continuing 
status of inoperable functions through the use of Station Log 
entries and Plant Status board entries. In addition, during 
operation with an inoperable function in the untripped, bypassed 
condition, the remaining RPS channel key-lock channel bypass 
switches will be ``Hold Carded'' (tagged) to prevent their operation 
without prior management approval consistent with the requirements 
of TS Table 3.5.1-1. Plant management maintains the responsibility 
to approve continued operation with inoperable functions unbypassed 
in the untripped state to ensure that the plant is operated in the 
safest configuration with regard to the extent of the failure, and 
the plant operating conditions. Prior to placing any channel of RPS 
or EFIC in bypass, the operator checks the status of redundant 
systems for operability and TS compliance and takes the proper 
action as required by existing plant conditions, plant operating 
procedures and TS. Should a channel of RPS contain an inoperable 
function unbypassed in the untripped condition which does not affect 
an EFIC channel, any channel of EFIC may be placed in maintenance 
bypass. RPS and EFIC remain single-failure proof in this 
configuration.
    The clarification of TS 3.5.1.3 which directs the operator to 
the appropriate actions if multiple channels become inoperable, or 
in the event of an inoperable channel or inoperable function 
occurring concurrent with one channel in bypass is considered to be 
administrative in nature. The change to Note 6 or Table 3.5.1-1 
results in the correction of misleading information and directs the 
Operator to place the plant in a safe mode depending on the system 
which is affected by a failure, and is also considered to be 
administrative in nature. The Bases changes add additional 
information to clarify the 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., et al., Docket No. 50-416, Grand Gulf 
Nuclear Station, Unit 1, Claiborne County, Mississippi

    Date of amendment request: January 13, 1994
    Description of amendment request: This amendment requests the 
removal of the interim technical specification limit on the number of 
spent fuel assemblies that may be stored in the spent fuel pool at 
Grand Gulf Nuclear Station.
    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. No significant increase in the probability or consequences of 
an accident previously evaluated results from this change.
    The NRC approved the installation of high density spent fuel 
storage racks in Amendment 17 to the Grand Gulf Nuclear Station 
(GGNS) Operating License. This amendment also brought GGNS into 
compliance with Standard Review Plan criteria which required 
maintaining the spent fuel pool at less than or equal 140 deg.F. The 
140 deg.F Technical Specifications (TS) limit remains in effect 
thereby preventing operation at excessive temperatures.
    The only outstanding question from Amendment 17, which resulted 
in a 2324 assembly technical specification limit, was whether the 
fuel pool cooling system could handle the heat load of a full fuel 
pool without excessive reliance on residual heat removal for 
extensive fuel pool cooling assist. Entergy Operations' proposed 
solution to this question was accepted in the NRC's letter dated 
July 30, 1992. The NRC accepted the solution pending submittal of 
results from tests to verify the specified flows. These results were 
submitted in a letter dated November 08, 1993.
    With previous approval of the installation of the high density 
spent fuel storage racks, the confirmation of adequate heat removal 
capability, and the 140 deg.F TS temperature limit, removal of the 
2324 limit to allow full use of the spent fuel pool would not cause 
an increase in the probability or consequences of an accident 
previously evaluated.
    2. This change would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The additional heat load generated by a full spent fuel pool 
(4348 assemblies) was evaluated. The evaluation concluded that full 
use of the spent fuel pool storage spaces would not exceed the 
temperature limits as are currently in place with the 2324 limit. 
The NRC letter dated July 30, 1992 and Entergy Operations letter 
dated November 08, 1993 resolved all outstanding heat removal 
questions. Therefore, this change would not create the possibility 
of a new or different kind of accident from any previously analyzed.
    3. This change would not involve a significant reduction in a 
margin of safety.
    Entergy Operations demonstrated in their November 01, 1991 
letter that the fuel pool temperature could be maintained at or 
below 140*F as specified in TS 3/4.7.9. This letter also 
demonstrated the ability to handle single active failures. Approval 
of measures outlined in this letter was provided in a Safety 
Evaluation Report contained in an NRC letter dated July 30, 1992.
    Given the 140 deg.F maximum temperature requirement as contained 
in TS 3/4.7.9 and compliance with single active failure criteria, 
this change would not involve a significant reduction in a margin of 
safety.
    Based on the above evaluation in accordance with 10CFR50l92(c), 
Entergy Operations, Inc. has concluded that operation in accordance 
with the proposed amendment involves no significant hazards 
considerations.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Judge George W. Armstrong 
Library, Post Office Box 1406, S. Commerce at Washington, Natchez, 
Mississippi 39120
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point Plant Units 3 and 4, Dade County, Florida

    Date of amendment request: December 28, 1993
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) for Turkey Point Units 3 and 4 
to incorporate features for steam generator (SG) overfill protection. 
Specifically, TS Tables 3.3-2, 3.3-3, 4.3-2 and the associated BASES 
section would be revised to add SG Water Level-High-High protection 
logic, instrumentation trip setpoints and surveillance requirements. 
The proposed TS changes would be in accordance with NRC Generic Letter 
(GL) 89-19, ``Safety Implication of Control Systems in LWR Nuclear 
Power Plants.''
    Basis for proposed no significant hazards consideration 
determination: As a result of the technical resolution of USI A-47, 
``Safety Implications of Control Systems in LWR Nuclear Power Plants,'' 
the Nuclear Regulatory Commission (NRC or the staff) concluded that all 
Pressurized Water Reactors (PWR) plants should provide automatic SG 
overfill protection. On September 20, 1989, the staff issued GL 89-19 
and recommended that plant procedures and TS include provisions for 
automatic SG overfill protection including surveillance requirements to 
assure that automatic overfill protection is available to mitigate main 
feedwater overfeed events during reactor power operation.
    The licensee proposed TS changes in response to GL 89-19. No 
physical changes to the plant would be required as a result of the 
proposed license amendments.
    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:
    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Including the SG Overfill protection requirements in the 
Technical Specifications is not assumed in the initiation of any 
analyzed event. These amendments will not increase the probability 
or consequences of an accident previously evaluated since the SG 
overfill event is not required or assumed for accident mitigation in 
any Updated Final Safety Analysis Report (UFSAR) safety analyses 
that comprise Turkey Point licensing basis. The additional 
requirements for the SG overfill system helps ensure that continuous 
addition of feedwater and carryover of excessive moisture to the 
turbine, is prevented. As a result, equipment protection is improved 
by the availability of this system function. As such, operation of 
the facility in accordance with the proposed amendments would not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The operation of the facility will not change as a result of the 
proposed license amendments, since Turkey Point currently maintains 
this protection logic. This change involves only the inclusion of 
the systems requirements into the Technical Specifications. The 
proposed change will not impose any new or unique requirements. 
Therefore, operation of the facility in accordance with the proposed 
amendments will 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 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed change does not involve a significant reduction in 
a margin of safety as the function, operation and testing of the 
installed SG Overfill protection is not described in the UFSAR. In 
addition, the SG overfill protection logic is not required or 
assumed for accident mitigation in any of the safety analyses that 
comprise the Turkey Point licensing basis. The proposed change 
formalizes the existing design, operating and testing requirements 
in the Technical Specifications. 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 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199
    Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer, 
P.C., 1615 L Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, 
Burke County, Georgia

    Date of amendment request: December 30, 1993
    Description of amendment request: The proposed change would allow a 
one time extension of the allowable outage time for each residual heat 
removal (RHR) pump from 3 to 7 days to allow modifications to the RHR 
system while the plant is in Mode 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change to the Technical Specifications does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated because the redundant train will 
remain available to assure that the RHR will respond to an accident 
as assumed in the accident analysis. A one time increase in the 
allowable outage time from 3 to 7 days has been shown to have only a 
small effect on the calculated frequency of core damage.
    2. The proposed change to the Technical Specifications does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated because the change only results in 
a one time increase of the allowable outage time. It does not result 
in an operational condition different from that which has already 
been considered by the Technical Specifications.
    3. The proposed addition to the Technical Specifications does 
not involve a significant reduction in a margin of safety because 
the effects of increasing the allowed outage time on the calculated 
core damage frequency has been evaluated and determined to be small.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308
    NRC Project Director: Loren R. Plisco, Acting

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, 
Linn County, Iowa

    Date of amendment request: December 22, 1993
    Description of amendment request: The proposed amendment would make 
editorial changes to correct typographical and administrative errors in 
the Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1) The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The amendment would only correct 
administrative and typographical errors. No physical changes to the 
plant or to the operation of the plant would result from this 
amendment.
    2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any evaluated previously. The 
amendment would only correct administrative and typographical 
errors. No physical changes to the plant or to operation of the 
plant would result from this amendment.
    3) The proposed amendment will not reduce the margin of safety. 
The amendment would only correct administrative and typographical 
errors. No physical changes to the plant or to operation of the 
plant would result from this amendment.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, 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: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401
    Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea, 
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC 
20036
    NRC Project Director: John N. Hannon

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, 
Linn County, Iowa

    Date of amendment request: January 21, 1994
    Description of amendment request: The proposed amendment would 
change the name of the company licensed to own a share of and operate 
the Duane Arnold Energy Center (DAEC) from Iowa Electric Light and 
Power Company to IES Utilities Incorporated, wherever it is referenced 
in the Operating License and Technical Specifications for DAEC. The 
title of the position responsible for the management of the Nuclear 
Division has also been changed to ``Vice President, Nuclear'' from 
``Manager-Nuclear Division.''
    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. No physical or operational changes to the DAEC 
will result from changing the corporate name or the position title. 
The DAEC will continue to be operated in the same manner with the 
same organization. The position title change results from the 
elimination of a layer of management. Formerly, the Manager-Nuclear 
Division reported through the Vice President, Production to the 
President of IELP. Now the Nuclear Division is headed by the Vice 
President, Nuclear who reports directly to the President of the 
corporation.
    2) The proposed amendment does not create the possibility of a 
new or different kind of accident from any previously evaluated. No 
physical or operational changes will result. The title change 
results from the elimination of a layer of management.
    3) The proposed change will not reduce any margin of safety. 
This change only revises the operating company name and changes a 
title.
    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: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401
    Attorney for licensee: Jack Newman, Esquire, Kathleen H. Shea, 
Esquire, Newman and Holtzinger, 1615 L Street, NW., Washington, DC 
20036
    NRC Project Director: John N. Hannon

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of amendment requests: January 17, 1994
    Description of amendment requests: The proposed amendments would 
change Technical Specification (TS) 3/4.1.3 for both units to increase 
the limit for control rod misalignment at or below 85% rated thermal 
power (RTP). The proposed changes would also increase the TS limit for 
control rod 8misalignment about 85% RTP if there is sufficient margin 
in the heat flux (FQ(Z)) and the nuclear enthalpy 
(FNdelta H) hot channel factors.
    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:
    Per 10 CFR 50.92, a proposed amendment to an operating license 
will not involve a significant hazards consideration if the proposed 
amendment satisfies the following three criteria:
    1. Does not involve a significant increase in the probability or 
consequences of an accident previously analyzed,
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed or evaluated, or
    3. Does not involve a significant reduction in a margin of 
safety.
    Criteria 1 and 3
    As seen in Attachment 4 [of the amendment request], sufficient 
margin exists in power distribution at 85% RTP to allow for 
increased misalignment. At 100% RTP, increased misalignment is 
allowed only if there is adequate margin in the peaking factors. 
Therefore, initial conditions remain unchanged from that assumed in 
the safety analyses. As far as the dropped rod and rod ejection 
accidents are concerned, the analyses were performed with 
conservative assumptions to envelope the increased misalignment. It 
should be noted that the power dependent insertion limit for Unit 1 
will be changed in a conservative manner at the beginning of cycle 
14. Based on these analyses, it is concluded that the proposed T/S 
changes do not significantly increase the probability or 
consequences of a previously analyzed accident or constitute a 
significant reduction in the margin of safety.
    Criterion 2
    The proposed T/S changes will not result in physical changes to 
the plant. Therefore, we believe that the proposed T/S changes will 
not create the possibility of a new or different kind of accident 
from any previously evaluated. Also, operation of the reactor with 
possible deeper rod insertion will not create the possibility of a 
new or different kind of accident.
    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 requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: A. Randolph Blough, Acting

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1, Oswego County, New York

    Date of amendment request: January 21, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.6.3, (Emergency Power Sources) to 
eliminate unnecessary testing of an operable emergency diesel generator 
(EDG) when the redundant EDG becomes inoperable. Eliminating 
unnecessary testing will potentially increase EDG reliability by 
reducing the stresses caused by such testing. The licensee stated that 
this proposed change is consistent with the guidance provided in NUREG-
1366, ``Improvements to Technical Specifications Surveillance 
Requirements,'' and NUREG-1433, ``Improved Standard Technical 
Specifications, General Electric Plants.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Technical Specification 4.6.3.e requires that the operable 
diesel-generator be manually started and operated at rated load for 
a minimum time of one hour immediately and once per week thereafter 
in the event any diesel-generator becomes inoperable.
    Niagara Mohawk proposes to revise Technical Specification 
4.6.3.e such that if a diesel-generator is declared inoperable due 
to preplanned maintenance or testing or due to a support system 
being inoperable, redundant diesel-generator testing would not be 
required. Declaring a diesel-generator inoperable due to preplanned 
maintenance or testing or due to a support system being inoperable 
does not affect the reliability of the operable diesel-generator nor 
does it in any way imply that a common cause failure exists.
    The normally required Technical Specification surveillance 
testing schedule demonstrates acceptable reliability and assures 
that the operable diesel-generator is capable of performing its 
intended safety function.
    Niagara Mohawk proposes to add wording to Technical 
Specification 4.6.3.e to permit an operator to evaluate a diesel-
generator failure to determine if a common cause failure exists 
before requiring testing of the redundant diesel-generator. As noted 
above, the intent of the additional diesel-generator testing is, in 
part, to determine if a common cause failure exists. Once the 
potential for a common cause failure has been examined and 
dismissed, testing beyond the normal surveillance schedule is 
excessive and does not contribute to improved diesel-generator 
reliability. Within eight (8) hours, the determination that no 
common cause failure exists is required to be completed or the 
operable diesel-generator will be tested. Eight (8) hours is 
consistent with the guidance provided in NUREG-1366, ``Improvements 
to Technical Specifications Surveillance Requirements.''
    Technical Specification 4.6.3.e requires that the operable 
diesel-generator be operated at rated load (i.e., connected to 
offsite power) to demonstrate its operability in the event any 
diesel-generator becomes inoperable. As indicated in Information 
Notice 84-69, when a diesel-generator is operated connected to 
offsite or non-vital loads, the emergency power system is not 
independent of disturbances on the non-vital and offsite power 
systems. Therefore, diesel-generator availability is potentially 
lessened by a demonstration of operability requiring connection of 
the diesel-generator to offsite power sources. At a time when at 
least one diesel-generator is already inoperable, this Surveillance 
Requirement could add further risk to losing the remaining operable 
diesel-generator. Therefore, Niagara Mohawk proposes that 
Surveillance Requirement 4.6.3.e be changed such that a diesel-
generator does not have to be operated at rated load. These changes 
will preclude offsite power source disturbances from affecting 
diesel-generator reliability.
    Existing Technical Specification 4.6.3.e requires that the 
operable diesel-generator be started immediately in the event a 
diesel-generator becomes inoperable. The requirement to immediately 
test a diesel-generator is overly burdensome when compared to more 
recent diesel-generator Technical Specification requirements. As 
previously discussed, Niagara Mohawk proposes to add wording to 
Technical Specification 4.6.3.e to give an operator eight (8) hours 
to determine whether a common cause failure exists or to test the 
operable diesel-generator when a diesel-generator is declared 
inoperable for a reason other than an inoperable support system or 
preplanned maintenance or testing. Eight (8) hours is consistent 
with the guidance provided in NUREG-1366, ``Improvements to 
Technical Specifications Surveillance Requirements.''
    Existing Technical Specification 4.6.3.e requires that the 
operable diesel-generator be tested immediately and once per week 
thereafter. Technical Specification 3.6.3.c requires that an 
inoperable diesel-generator be returned to an operable condition 
within seven (7) days to meet the Limiting Condition for Operation. 
Therefore, the requirement to test the operable diesel-generator 
``once a week thereafter'' is not applicable. In addition, testing 
the operable diesel-generator one time is adequate to confirm 
operability of a diesel-generator. Repetitive testing following 
initial confirmation of operability is unwarranted. Therefore, 
Niagara Mohawk proposes to delete the requirement to test the 
operable diesel-generator weekly following the initial test.
    Because the proposed change does not affect the design or 
performance of the diesel-generators nor adversely affect the 
reliability of the diesel-generators, the change will not result in 
an increase in the consequences of an accident previously evaluated 
(i.e., Station Blackout analyses). Because this change does not 
affect the probability of accident precursors, the proposed change 
does not affect the probability of an accident previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated
    The proposed change to Technical Specification 4.6.3.e does not 
introduce any new operating configurations or new accident 
precursors and does not involve any physical alterations to plant 
configurations which could initiate a new or different kind of 
accident. The proposed change does not affect the design or 
performance characteristics of the diesel-generators nor does the 
change create the possibility of the loss of both diesel-generators 
because common cause failure assessments will be performed. The 
change will delete excessive diesel-generator testing and therefore 
increase overall plant safety by increasing diesel-generator 
reliability. Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety
    The proposed change to Technical Specification 4.6.3.e will not 
reduce the number of emergency power sources required by Technical 
Specification Limiting Condition for Operation 3.6.3 or affect the 
normal surveillance requirements as described in Technical 
Specification 4.6.3. The normal surveillance tests demonstrate 
acceptable reliability and assure that the operable diesel-generator 
is capable of performing its intended function. The proposed change 
to delete the excessive testing requirements does not affect the 
design or performance of any diesel-generator and does not adversely 
affect diesel-generator reliability. Eliminating unnecessary testing 
will potentially increase diesel-generator reliability by reducing 
the stresses caused by such testing. 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 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Robert A. Capra

Northern States Power Company, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota

    Date of amendment request: November 30, 1993
    Description of amendment request: The proposed amendment would 
change sections 3.2/4.2, Protective Instrumentation, and 3.17/4.17, 
Control Room Habitability, by deleting the requirements for a chlorine 
detection system and revises the limiting conditions for operation for 
the Control Room Ventilation System to be more consistent with Standard 
Technical Specifications. Due to design changes at the Monticello 
Nuclear Generating Plant, chlorine is no longer stored onsite as a 
liquified gas and regulations requiring early warning of an onsite 
chlorine release do not apply.
    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. Concerning Deletion of Requirements for the Chlorine 
Detection System
    The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Postulated chemical releases of chlorine have been shown to be 
such that incapacitation of the control room operators would not 
occur within allowed time frames for the donning of protective 
breathing equipment, or that the probability of a chlorine trucking 
transportation accident which causes incapacitation of control room 
operators with potential consequences of a radioactive release in 
excess of 10 CFR 100 guidelines is well below the level of concern 
as established in regulatory guidance. Therefore, this amendment 
will not cause a significant increase in the probability or 
consequences of an accident previously evaluated for the Monticello 
plant.
    The proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously analyzed.
    The performance of a new toxic chemical analysis for the 
Monticello site has demonstrated that human detection may be relied 
upon to detect chlorine toxic chemical releases. Operator protection 
is established via the donning of protective breathing equipment. 
The capability to manually isolate the control room with dampers is 
retained. The ability of the operators to cope with a chlorine toxic 
gas hazard remains consistent with the protection measures available 
for other toxic chemicals stored onsite, stored in the vicinity of 
the site, or transported near the plant site. The proposed amendment 
will not create the possibility of a new or different kind of 
accident.
    The proposed amendment will not involve a significant reduction 
in the margin of safety.
    The performance of a new toxic chemical analysis for the 
Monticello site has demonstrated that incapacitation of the control 
room operators would not occur within allowed time frames for the 
donning of protective breathing equipment and that a postulated 
hazardous chemical release due to a trucking transportation accident 
involving chlorine is of a sufficiently low probability of 
occurrence that it need not be considered. The basis of the chlorine 
detectors and associated Technical Specifications is to provide 
protection against an accident scenario which has been demonstrated 
to be of extremely low probability (a trucking transportation 
accident involving chlorine within five miles of the plant), 
therefore removal of the chlorine detectors from the plant design 
and the associated Technical Specifications will not involve a 
significant reduction in the margin of safety.
    2. Concerning the Limiting Conditions for Operation for the 
Control Room Ventilation System and Technical Specification Bases

    The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The Control Room Ventilation system ensures that Main Control 
Room habitability is maintained such that personnel and equipment 
located in the control room can respond to mitigate the consequences 
of an accident. The system does not contribute to the probability of 
occurrence of any design basis accident. The operability 
requirements as proposed for the revised specification 3.17.A ensure 
that the Control Room Ventilation system is operable during plant 
conditions for which significant radioactive releases are postulated 
consistent with the Standard Technical Specification. The proposed 
changes ensure the Control Room Ventilation system is restored to an 
operable status or that actions are taken to minimize the importance 
of the system function within time frames which take into 
consideration the low probability of an event occurring which would 
require Control Room Ventilation system function. Therefore, this 
amendment will not cause a significant increase in the probability 
or consequences of an accident previously evaluated for the 
Monticello plant.
    The proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously analyzed.
    The proposed changes to Technical Specifications 3.17.A do not 
alter the function of the Control Room Ventilation system or its 
interrelationships with other systems. The proposed changes provide 
requirements to ensure the Control Room Ventilation system is 
capable of performing its required function or that actions are 
taken to minimize the potential for its function being required 
consistent with regulatory guidance; therefore, this amendment will 
not create the possibility of a new or different kind of accident 
from any accident previously analyzed.
    The proposed amendment will not involve a significant reduction 
in the margin of safety.
    The operability requirements as proposed for the revised 
specification 3.17.A ensure that the Control Room Ventilation system 
is operable during plant conditions for which significant 
radioactive releases are postulated. The performance of a new toxic 
chemical analysis for the Monticello site has demonstrated that a 
postulated hazardous chemical release due to a trucking 
transportation accident involving chlorine is of a sufficiently low 
probability of occurrence that it need not be considered. As the 
basis of the chlorine detectors and current operability requirements 
for the control Room Ventilation system is to provide protection 
against an accident scenario which has been demonstrated to be of 
extremely low probability, the proposed revision to the Control Room 
Ventilation operability requirements will not involve a significant 
reduction in the margin of safety.
    The proposed changes to Technical Specification 3.17.A ensure 
that both trains of the Control Room Ventilation system are restored 
to an operable status within a time frame which takes into 
consideration the low probability of an event occurring requiring 
Control Room Ventilation system function, the availability of the 
redundant Control Room Ventilation train and the capability of the 
safety related Emergency Filtration Train to pressurize the control 
room without the Control Room Ventilation system. The proposed 
changes provide requirements to ensure the Control Room Ventilation 
system is capable of performing its required function or that 
actions are taken to minimize the potential for its function to be 
required consistent with regulatory guidance; therefore, the 
proposed change will 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: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: L. B. Marsh

Northern States Power Company, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota

    Date of amendment request: January 3, 1994
    Description of amendment request: The proposed amendment would 
revise the requirements of Technical Specification 4.6.E.1.a, which 
currently specifies that a minimum of seven safety/relief valves shall 
be bench checked or replaced with a bench checked valve each refueling 
outage. The proposed amendment would change this specification to 
require the valves to be tested in accordance with the Section XI 
Inservice Testing Requirements of the ASME Boiler and Pressure Vessel 
Code. The proposed change is consistent with the Improved Standard 
Technical Specifications, NUREG-1433.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    a. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment is limited to changes to the surveillance 
testing requirements (bench checking or replacement) applicable to 
the main steam system safety/relief valves. This surveillance 
requirement is performed while the plant is in a cold shutdown 
condition at a time when the safety/relief valves are not required 
to be operable. The performance of this evolution is not an input or 
consideration in any accident previously evaluated, thus the 
proposed change will not increase the probability of any such 
accident occurring. Current safety analyses conclude that the 
pressure relief capabilities of the Safety Relief valves are 
adequate assuming that one of the eight safety/relief valves fails 
to open upon demand. The proposed change will not adversely affect 
the reliability of the valves and will therefore not reduce the 
conservatism of this assumption.
    Similarly, the proposed amendment specifies testing requirements 
consistent with accepted industry codes and regulatory guidance to 
provide assurance that the valves will function as designed. The 
amendment will not diminish the capability of the safety/relief 
valves to perform as required during any accident previously 
evaluated and will therefore not increase the consequences of any 
such accident.
    b. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed amendment does not involve any modification to 
plant equipment or operating procedures, nor will it introduce any 
new safety/relief valve failure modes that have not been previously 
considered. The net result of the proposed amendment will be to 
allow the plant staff the option of decreasing the frequency of 
safety/relief valve testing to a level that has been acknowledged as 
acceptable by the ASME Code and NUREG-1433. We therefore conclude 
the proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously analyzed.
    c. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed amendment does not involve a decrease in the number 
or capacity of safety/relief valves that are provided in the system, 
nor does it involve any change in safety/relief valve setpoints, 
operability requirements, or limiting conditions for operation. 
Based on these considerations, we conclude the proposed amendment 
will 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: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: L. B. Marsh

Northern States Power Company, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota

    Date of amendment request: January 4, 1994
    Description of amendment request: The proposed amendment would 
change Technical Specifications section 3.11, Reactor Fuel Assemblies, 
by removing information concerning the analytical method to determine 
average planar linear heat generation rate (APLHGR) and providing 
reference to the presentation of the information in the Core Operating 
Limits Report. In addition, this proposed amendment would change 
section 6.7, Reporting Requirements, by revising the listing of 
approved analytical methods for developing the Core Operating Limits 
Report, and it would revise the Technical Specification Bases for 
section 3.11 concerning the calculation methodology for MCPR [minimum 
critical power ratio]. The proposed change to specification 3.11.A 
would eliminate the duplication of requirements specified in 
specification 6.7.A.7 and the Core Operating Limits Report for 
establishing APLHGR limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The APLHGR limits originate from and are associated with LOCA 
[loss-of-coolant accident] analyses. Standard exposure dependent 
APLHGR limits are generated from LOCA analyses initiated from rated 
power and flow conditions. For any allowable off power and off flow 
condition the APLHGR limit is the smaller of the flow dependent or 
power dependent limit. These limits are also used in the fuel 
thermal-mechanical analysis and transient analysis. Flow dependent 
APLHGR requirements will continue to be established based on 
analysis and fuel type specific limits determined using NRC approved 
methodologies to ensure that peak transient average planar heat 
generation rate during these events is not increased above the fuel 
design basis values. Power dependent APLHGR limits will continue to 
be established based on analysis and fuel type specific limits 
determined using NRC approved methodologies to ensure that peak 
transient average planar heat generation rate during any transient 
is not increased above the rated fuel design basis transient values. 
The proposed amendment establishes appropriate controls to ensure 
that the APLHGR limits will continue to be determined and 
established using NRC approved methodology; therefore, this 
amendment will not cause a significant increase in the probability 
or consequences of an accident previously evaluated for the 
Monticello plant.
    The proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously analyzed.
    The proposed amendment does not involve any modifications to 
plant equipment or operating procedures, nor will it introduce any 
new failure modes. The proposed amendment ensures that cycle 
specific APLHGR limits are determined and established using approved 
methodologies and will not create the possibility of a new or 
different kind of accident.
    The proposed amendment will not involve a significant reduction 
in the margin of safety.
    The proposed amendment removes duplication which exists in the 
Monticello Technical Specification for the identification of the 
approved analytical methods for establishing the APLHGR core 
operating limit. In addition the proposed amendment adds the NRC 
approved Siemens' analytical method for the determination of APLHGR 
limits based on LOCA/ECCS [emergency core cooling system] analyses. 
Inclusion of the NRC approved Siemens' analytical method ensures 
proper coordination of the methodology employed to establish the 
APLHGR limiting condition for operation for each type of fuel as a 
function of axial location and average planar exposure. APLHGR 
limits will continue to be determined using NRC approved methodology 
as established in specification 6.7.A.7.b. The established APLHGR 
limits will be verified to be consistent with the accident analysis 
contained in the Monticello Updated Safety Analysis Report. The 
proposed amendment will 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: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: L. B. Marsh

Northern States Power Company, Docket Nos. 50-282 and 50-306, 
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue 
County, Minnesota

    Date of amendments request: September 21, 1992, as revised December 
29, 1992, and November 24, 1993
    Description of amendments requests: The proposed amendments would 
change various Technical Specification (TS) sections and associated 
Bases for surveillance test intervals and allowed outage times for the 
engineered safety features and reactor protection system 
instrumentation consistent with the NRC Staff position as documented in 
NRC letters to the Westinghouse Owners Group.
    The proposed license amendment request also updates operation modes 
to be consistent with Westinghouse Standard Technical Specification 
operational modes and also includes several editorial changes to the 
Prairie Island TS that are unrelated to the changes described above.
    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 will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The determination that the results of the proposed change are 
within all acceptable criteria have been established in the SERs 
prepared for WCAP-10271, WCAP-10271 Supplement 1, WCAP-10271 
Supplement 2 and WCAP-10271 Supplement 2, Revision 1 issued by 
References 1, 2, and 5 [of the November 24, 1993, application]. 
Implementation of the proposed changes is expected to result in an 
acceptable increase in total Reactor Protection and Engineered 
Safety Features Systems yearly unavailability. This increase, which 
is primarily due to less frequent surveillance, results in a[n] 
increase of similar magnitude in the probability of an Anticipated 
Transient Without Scram (ATWS) and in the probability of core melt 
resulting from an ATWS and also results in a small increase in core 
damage frequency (CD) due to Engineered Safety Features 
unavailability.
    Implementation of the proposed changes is expected to result in 
a significant reduction in the probability of core melt from 
inadvertent reactor trips. This is a result of a reduction in the 
number of inadvertent reactor trips (0.5 fewer inadvertent reactor 
trips per unit per year) occurring during testing of Reactor 
Protection System instrumentation. This reduction is primarily 
attributable to less frequent surveillance.
    The reduction in inadvertent core melt frequency is sufficiently 
large to counter the increase in ATWS core melt probability 
resulting in an overall reduction in total core melt probability.
    The values determined by the Westinghouse Owners Group and 
presented in the WCAP for the increase in core damage frequency were 
verified by Brookhaven National Laboratory (BNL) as part of an audit 
and sensitivity analyses for the NRC Staff. Based on the small value 
of the increase compared to the range of uncertainty in the core 
damage frequency, the increase is considered acceptable.
    The changes of an editorial nature, including the change to 
Standard Technical Specification format for the instrumentation 
Technical Specifications and mode definitions, have no impact on the 
severity or consequences of an accident previously evaluated.
    The proposed changes do not result in an increase in the 
severity or consequences of an accident previously evaluated. 
Implementation of the proposed changes affects the probability of 
failure of the Reactor Protection System and Engineered Safety 
Features but does not alter the manner in which protection is 
afforded nor the manner in which limiting criteria are established.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed changes do not involve hardware changes and do not 
result in a change in the manner in which the Reactor Protection 
System and Engineered Safety Features provide plant protection. No 
change is being made which alters the functioning of the Reactor 
Protection System or Engineered Safety Features. Rather the 
likelihood or probability of the Reactor Protection System or 
Engineered Safety Features functioning properly is affected as 
described above. Therefore the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The changes of an editorial nature, including the change to 
Standard Technical Specification format for the instrumentation 
Technical Specifications and mode definitions does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    . The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system setpoints or limiting conditions for 
operation are determined. The impact of reduced testing other than 
as addressed above is to allow a longer time interval over which 
instrument uncertainties (e.g., drift) may act. Experience has shown 
that the initial uncertainty assumptions are valid for reduced 
testing.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety by:
    a. Less frequent testing will result in less inadvertent reactor 
trips and actuation of Engineered Safety Features components.
    b. Higher quality repairs leading to improved equipment 
reliability due to longer repair times.
    c. Improvements in the effectiveness of the operating staff in 
monitoring and controlling plant operation. This is due to less 
frequent distraction of the operator and shift supervisor to attend 
to instrumentation testing.
    The changes of an editorial nature, including the change to 
Standard Technical Specification format for the instrumentation 
Technical Specifications and mode definitions [do] not lead to a 
reduction in any margin of safety.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: L. B. Marsh

Omaha Public Power District, Docket No. 50-285, Fort Calhoun 
Station,Unit No. 1, Washington County, Nebraska

    Date of amendment request: December 28, 1993
    Description of amendment request: The proposed amendment to the 
Technical Specifications would revise the surveillance test frequency 
from monthly to quarterly for several channel functional tests for 
Reactor Protective System and Engineered Safety Feature Instrumentation 
and Controls based on Generic Letter 93-05.
    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 significant hazards 
considerations because operation of Fort Calhoun Station Unit (FCS) 
No. 1 in accordance with this change would not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Increasing the surveillance test interval (STI) from monthly to 
quarterly for the Reactor Protective System (RPS) and Engineered 
Safety Features Actuation System (ESFAS) instrumentation has two 
principal effects with opposing impacts on core melt risk. The first 
impact is a slight increase in core melt frequency that results from 
the increased unavailability of the instrumentation in question. The 
unavailability of the tested instrumentation components is 
translated to result in a failure of the reactor to trip, an 
Anticipated Transient Without Scram (ATWS), or a failure of the 
appropriate engineered safety features to actuate when required. The 
opposing impact on core melt risk is the corresponding reduction in 
core melt frequency that would result due to the reduced exposure of 
the plant to test-induced transients. This results in a net decrease 
in core melt frequency of approximately 4.1x10-8 per year.
    Representative fault tree models were developed for FCS and the 
corresponding changes in core melt frequency were quantified in 
evaluations CEN-327-A and CEN-327-A, Supplement 1. The NRC issued a 
Safety Evaluation Report (SER) which found that these evaluations 
were acceptable for justifying the extensions in the STIs for the 
RPS and ESFAS from 30 days to 90 days and that the RPS 
unavailabilities resulting from extending the STIs were not 
considered to be significant. Estimates of the reduction in scram 
frequency from the reduction in test-induced scrams and the 
corresponding reduction in core melt frequency were found 
acceptable. STIs of 90 days were found to result in a net reduction 
in core melt risk.
    A plant specific calculation/setpoint drift analysis was 
conducted, as required by the NRC SER, that analyzed the effect on 
instrument drift of extending the RPS and ESF instrumentation and 
controls functional STI from monthly to quarterly. The results 
demonstrated that the observed changes in instrument uncertainties 
for the extended STI do not exceed the current 30-day setpoint 
assumptions. Therefore, it is unnecessary to change any setpoints to 
accommodate the proposed extended STI.
    Operation of the facility in accordance with this proposed 
change, therefore, will 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.
    The proposed change does not involve any changes in equipment 
and will not alter the manner in which the plant will be operated. 
RPS and ESFAS setpoints will not be changed as the instrument 
uncertainties resulting from the proposed STI (calculated using 
actual plant data) are less than the instrument uncertainties 
assumed for 30 days. Thus, this proposed change will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) Involve a significant reduction in a margin of safety.
    There are no changes to the equipment or plant operations. RPS 
and ESFAS setpoints will not be changed as the instrument 
uncertainties resulting from the proposed STI (calculated using 
actual plant data) are less that the instrument uncertainties 
assumed for 30 days.
    Implementation of the proposed changes is expected to result in 
an overall improvement in plant safety due to the fact that reduced 
testing intervals will result in fewer inadvertent reactor trips and 
less frequent actuation of ESFAS components. The conclusions of the 
accident analyses in the FCS Updated Safety Analysis Report (USAR) 
remain valid and the safety limits continue to be met. Thus, this 
proposed change does not reduce 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: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875 
Connecticut Avenue, N.W., Washington, D.C. 20009-5728NRC Project 
Director:
    William D. Beckner

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: December 28, 1993
    Description of amendment request: The proposed amendment to the 
James A. FitzPatrick Technical Specifications (TSs) clarifies Limiting 
Condition for Operation (LCO) 3.5.D.4. Amendment No. 179 to the TS 
added LCO 3.5.D.4 to permit hydrostatic and leakage testing at 
temperatures up to 300 deg.F without requiring certain equipment, 
including the automatic depressurization system (ADS), to be operable. 
However, LCO 3.5.D.4 can be mistakenly interpreted to require the ADS 
be operable at temperatures less than 212 deg.F. Requiring the ADS to 
be operable during hydrostatic and leakage testing with temperatures 
below 212 deg.F was clearly not the intent of Amendment No. 179. The 
proposed change will clarify LCO 3.5.D.4 to resolve this concern.
    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 plant accident analyses are not affected by the proposed 
Technical Specification change. Prior to implementation of Amendment 
179, hydrostatic and leakage testing of the RCS was performed with 
reactor coolant temperatures below 212 deg.F while the ADS was 
inoperable. Amendment 179 revised the Technical Specifications in 
anticipation of increased pressure temperature limits requiring 
hydrostatic and leakage testing at or above 212 deg.F. Requiring the 
ADS to be operable during hydrostatic or leakage testing with 
temperatures below 212 deg.F was clearly not the intent of Amendment 
179. The change will not increase the probability or consequences of 
previously evaluated accidents.
    2. create the possibility of a new or different kind of accident 
from those previously evaluated.
    The proposed change involves no modifications to hardware, 
analyses, operations or procedures. The change clarifies LCO 3.5.D.4 
to allow hydrostatic and leakage testing of the RCS below 300 deg.F 
without requiring the ADS to be operable. The change is 
administrative in nature since it only clarifies the intent of the 
Technical Specifications as agreed to with the NRC and cannot create 
a new or different kind of accident.
    3. involve a significant reduction in the margin of safety.
    The proposed change will not affect any plant safety margins. 
The existing plant accident analyses are not affected by the 
proposed change. This revision of LCO 3.5.D.4 is intended to clarify 
that the ADS is not required to be operable during hydrostatic or 
leakage testing of the RCS. This position is substantiated by the 
NRC safety evaluation for Amendment 179 which acknowledges that 
hydrostatic and leakage testing can not be performed without making 
the ADS, and other systems, inoperable.
    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: Robert A. Capra

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: January 31, 1994
    Description of amendment request: The proposed amendment to the 
James A. FitzPatrick Technical Specifications would revise the limiting 
conditions for operation (LCO), surveillance requirements, and Bases 
section for the main condenser steam jet air ejectors (SJAE). The 
proposed changes correct a typographical error, clarify the modes of 
operation during which the SJAE LCOs and surveillance requirements are 
applicable, revise the action required upon entering a SJAE LCO, and 
establish a threshold level below which there will be no requirement to 
perform grab samples and isotopic analyses of SJAE effluent.
    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 amendment revision involves no hardware changes, no 
changes to the operation of any systems or components and no changes 
to structures. The changes clarify the Technical Specifications by 
specifying the modes of operation during which the LCOs and 
Surveillance Requirements of Specification 3.5 are applicable. The 
changes also include specific guidance for the operators to prevent 
or minimize the release of radioactive gases to the environment. 
These changes can not cause an increase in the probability of, nor 
alter the consequences of, an accident previously evaluated.
    The establishment of a threshold below which grab samples are 
not required will alter procedures by allowing SJAE operation 
without grab samples to determine effluent content at low levels of 
radioactivity (i.e., less than 5,000 micro Ci/sec). This will not 
affect the monitoring system's ability to detect, alarm, and isolate 
the offgas system if the concentration of radioactive material in 
the effluence reaches the appropriate setpoint.
    The surveillance requirement for taking a grab sample after a 
greater than 50% increase in release rate is intended to assist 
operators in determining if there is any increase in fuel failure 
during steady state operations. This would assure that routine 
operational limits are maintained. The grab samples do not provide 
any automatic protective function (e.g., MSIV [main steam isolation 
valve] or Offgas System isolation) for mitigating an accident but 
provide radionuclide concentration data.
    The performance of SJAE effluent grab samples is not credited 
towards detecting nor mitigating any design basis accidents since 
spontaneous fuel failure is not a FSAR [Final Safety Analysis 
Report] accident initiator but a consequence of an accident. 
Therefore, the use of a 5,000 micro Ci/sec threshold, which is 
approximately 1% of the trip setpoint, would not alter the 
consequences or probabilities of established accident scenarios.
    2. create the possibility of a new or different kind of accident 
from those previously evaluated.
    The proposed changes provide improved clarity concerning 
applicability of the specifications and specific guidance for 
preventing/mitigating the release of radioactive gases to the 
environment. The proposed changes also provide guidance for limiting 
the number of unnecessary grab samples.
    These changes do not affect the manner in which the main 
condenser steam jet air ejector is operated. The proposed changes to 
the Technical Specifications reflect either established plant 
practice (i.e., applicable modes or mitigation procedures) or new 
surveillance guidelines to minimize unnecessary grab samples. In all 
cases, the proposed changes have no affect on any parameters which 
would be considered or used in an accident analysis. The changes, 
therefore, do not pose a safety issue different from those analyzed 
previously for the FSAR.
    3. involve a significant reduction in the margin of safety.
    The proposed changes to the Technical Specifications will not 
alter the intent of the surveillance requirement to monitor for the 
possibility of fuel failure. Considering the difference between the 
proposed threshold value and the current alarm setpoint, a reduction 
in grab samples during plant operation with low concentrations of 
radioactivity in the primary coolant will not affect any plant 
safety margins.
    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: Robert A. Capra

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: January 31, 1994
    Description of amendment request: The proposed amendment to the 
James A. FitzPatrick Technical Specifications would revise 
Specification 3.8 to adopt the Limiting Conditions for Operation (LCO) 
of Section 3/4.7.6, ``Sealed Source Contamination,'' as stated in 
NUREG-0123, ``Standard Technical Specifications for General Electric 
Boiling Water Reactors (BWR/5)'' (STS). In addition, the proposed 
change reformats Specifications 3.8 and 4.8 to make them consistent 
with the remainder of the FitzPatrick 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 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.
    Adopting the LCO described in the ``Sealed Source 
Contamination'' section of NUREG-0123 (STS) does not increase the 
probability or the consequences of an accident or malfunction of a 
safety-related structure, system, or component previously reviewed 
in the FSAR [Final Safety Analysis Report]. The proposed changes do 
not increase the probability of causing, either directly or 
indirectly an uncontrolled release of significant amounts of 
radiation. Deleting 10 CFR 30.71 as the basis for exempting sealed 
sources for the leak testing requirements removes a requirement that 
is redundant to other federal regulations requirements. The proposed 
changes to reformat Specifications 3.8 and 4.8 are administrative in 
nature and do not increase the probability or consequences of an 
accident previously evaluated in the FSAR. Therefore, the proposed 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not alter the radioactive materials 
controls established at the restricted area boundaries and do not 
increase the amount of radioactive materials on site. There are no 
modifications to safety systems as a result of the proposed changes. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated in the FSAR.
    3. involve a significant reduction in a margin of safety.
    Adopting the wording of the STS regarding the sealed sources 
limiting conditions for operations will not reduce the ability of 
the operators to detect a leaking sealed radioactive source. 
Established radiological controls (i.e., handling techniques and 
good health physics practices) implemented through plant procedures 
will ensure that the sealed sources will continue to be tested as 
required by the Technical Specifications and applicable regulations. 
The proposed changes do not alter the radioactive materials controls 
established at the restricted area boundary and do not increase the 
amount of radioactive materials on site. Therefore, the proposed 
changes do not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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: Robert A. Capra

Sacramento Municipal Utility District, Docket No. 50-312, Rancho 
Seco Nuclear Generating Station, Sacramento County, California

    Date of amendment request: December 9, 1993
    Description of amendment request: The proposed amendment would 
change the Rancho Seco Permanently Defueled Technical Specifications 
(PDTS) to implement and ensure consistency with the revisions in 10 CFR 
Part 20.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    A significant increase in the probability or 
consequences of an accident previously evaluated in the SAR (Safety 
Analysis Report) will not be created, because the proposed changes 
are editorial in nature, are designed to implement the 10 CFR Part 
20 regulations, and have no affect on any accidents evaluated in the 
Rancho Seco Defueled Safety Analysis Report (DSAR), i.e., the 
dropped fuel assembly accident, the loss of offsite power condition, 
or a radwaste tank rupture.
    PA-187 (Proposed Amendment) will not create the 
possibility of a new or different type of accident evaluated in the 
SAR, because the changes are editorial in nature, implement the new 
10 CFR Part 20 radiation protection regulations, and do not provide 
any new mechanisms by which an accident can occur.
    The proposed PDTS amendment will not involve a 
significant reduction in the margin of safety, because the District 
will continue to maintain the appropriate radiation protection 
controls, through implementation of the new 10 CFR Part 20 
regulations, that are necessary to ensure Rancho Seco continues to 
be operated safely from a personnel radiation exposure standpoint 
during the Permanently Defueled Mode.
    The NRC staff has reviewed the analysis of the licensee 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: Central Library, Government 
Documents 828 I Street, Sacramento, California 95814.
    Attorney for licensee: Dana Appling, Esquire, Sacramento Municipal 
Utility District, P.O. Box 15830, Sacramento, California 95852-1830
    NRC Project Director: Seymour H. Weiss

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear 
Plant, Unit 2, Hamilton County, Tennessee

    Date of amendment request: February 8, 1994 (TS 94-02)
    Description of amendment request: The proposed change would revise 
Operating License Condition 2.C.(17) to temporarily extend the 
surveillance interval for certain specified instruments from the normal 
18-month interval to a maximum of 28 months for 18-month surveillances 
and 46 months for the 3-year Containment fire hose hydrostatic 
surveillance test in order to prevent exceeding the allowable testing 
frequency prior to the refueling outage that has been rescheduled to 
start in July 1994.
    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:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change is temporary and allows a one-time extension 
of specific surveillance requirements (SRs) for Cycle 6 to allow 
surveillance testing to coincide with the sixth refueling outage. 
The proposed surveillance interval extension is short and will not 
cause a significant reduction in system reliability nor affect the 
ability of the systems to perform their design function. Current 
monitoring of plant conditions and continuation of the surveillance 
testing required during normal plant operation will continue to be 
performed to ensure conformance with TS operability requirements. 
Therefore, this 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 previously analyzed.
    Extending the surveillance interval for the performance of 
specific testing will not create the possibility of any new or 
different kind of accidents. No changes are required to any system 
configurations, plant equipment, or analyses. Therefore, this 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.
    Surveillance interval extensions will not impact any plant 
safety analyses since the assumptions used will remain unchanged. 
The safety limits assumed in the accident analyses and the design 
function of the equipment required to mitigate the consequences of 
any postulated accidents will not be changed since only the 
surveillance test interval is being extended. Historical performance 
generally indicates a high degree of reliability, and surveillance 
testing performed during normal plant operation will continue to be 
performed to verify proper performance. Therefore, the plant will be 
maintained within the analyzed limits, and the proposed extension 
will not significantly reduce the margin of safety.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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 amendment request: December 23, 1992
    Description of amendment request: The proposed amendment would 
revise TS 3/4 3.3.5 and its Bases adding testing requirements for 
transfer switches used to meet 10 CFR Part 50, Appendix R (Fire 
Protection) requirements and specifies a new special report requirement 
for TS 6.9.2.
    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, indicating that the proposed 
changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because none of the proposed changes 
are associated with the initiation of any design bases accident. The 
addition of Limiting Condition for Operation (LCO) 3.3.3.5.2 and 
Surveillance Requirement (SR) 4.3.3.5.2 to the Technical 
Specifications will require each control circuit and transfer switch 
that is required for a serious control room or cable spreading room 
fire to be operable during Modes 1, 2 and 3 and to be verified at 
least once per 18 months as capable of performing the intended 
function. New Action b will require restoration of an inoperable 
control circuit or transfer switch (required for a serious control 
room or cable spreading room fire) within 30 days or a Special 
Report submitted to the NRC pursuant to Specification 6.9.2 within 
the next 30 days. Surveillance testing procedures will be prepared, 
reviewed and approved in accordance with Technical Specification 
(TS) 6.5.3, Technical Review and Control, which will ensure an 
unreviewed safety question is not created. To support the addition 
of the new LCO, Action and SR, the existing LCO, Action and SR are 
proposed to be administratively re-numbered or re-lettered. The new 
Special Report requirement is proposed to be administratively added 
to TS 6.9.2.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no equipment, accident 
conditions, or assumptions are affected which could lead to 
significant increases in radiological consequences. The addition of 
LCO 3.3.3.5.2 and SR 4.3.3.5.2 to the Technical Specifications will 
require each control circuit and transfer switch that is required 
for a serious control room or cable spreading room fire to be 
operable during Modes 1, 2 and 3 and to be verified at least once 
per 18 months as capable of performing the intended function. New 
Action b will require restoration of an inoperable control circuit 
or transfer switch (required for a serious control room or cable 
spreading room fire) within 30 days or a Special Report submitted to 
the NRC pursuant to Specification 6.9.2 within the next 30 days. 
Surveillance testing procedures will be prepared, reviewed and 
approved in accordance with Technical Specification (TS) 6.5.3, 
which will ensure an unreviewed safety question is not created. To 
support the addition of a new LCO, Action and SR, the existing LCO, 
Action and SR are proposed to be administratively re-numbered or re-
lettered. The new Special Report requirement is proposed to be 
administratively added to TS 6.9.2.
    2a. Not create the possibility of a new kind of accident from 
any accident previously evaluated because no new accident initiators 
are introduced by the proposed changes. The addition of LCO 
3.3.3.5.2 and SR 4.3.3.5.2 to the Technical Specifications will 
require each control circuit and transfer switch that is required 
for a serious control room or cable spreading room fire to be 
operable during Modes 1, 2 and 3 and to be verified at least once 
per 18 months as capable of performing the intended function. New 
Action b will require restoration of an inoperable control circuit 
or transfer switch (required for a serious control room or cable 
spreading room fire) within 30 days or a Special Report submitted to 
the NRC pursuant to Specification 6.9.2 within the next 30 days. 
Surveillance testing procedures will be prepared, reviewed and 
approved in accordance with TS 6.5.3, which will ensure an 
unreviewed safety question is not created. To support the addition 
of the new LCO, Action and SR, the existing LCO, Action and SR are 
proposed to be administratively re-numbered or re-lettered. The new 
Special Report requirement is proposed to be administratively added 
to TS 6.9.2.
    2b. Not create the possibility of a different kind of accident 
from any accident previously evaluated because no different accident 
initiators are introduced by the proposed changes. The addition of 
LCO 3.3.3.5.2 and SR 4.3.3.5.2 to the Technical Specifications will 
require each control circuit and transfer switch that is required 
for a serious control room or cable spreading room fire to be 
operable during Modes 1, 2, and 3 and to be verified at least once 
per 18 months as capable of performing the intended function. New 
Action b will require restoration of an inoperable control circuit 
or transfer switch (required for a serious control room or cable 
spreading room fire) within 30 days or a Special Report submitted to 
the NRC pursuant to Specification 6.9.2 within the next 30 days. 
Surveillance testing procedures will be prepared, reviewed and 
approved in accordance with TS 6.5.3, which will ensure an 
unreviewed safety question is not created. To support the addition 
of the new LCO, Action and SR, the existing LCO, Action and SR are 
proposed to be administratively re-numbered or re-lettered. The new 
Special Report requirement is proposed to be administratively added 
to TS 6.9.2.
    3. Not involve a significant reduction in a margin of safety 
because these are not new or significant changes to the initial 
conditions contributing to accident severity or consequences, 
therefore, there are no significant reductions 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 Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, N.W., Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: September 24, 1993
    Description of amendment request: The proposed amendment would 
revise Technical Specifications to extend the reporting period of the 
Semiannual Radioactive Effluent Release Report from semiannually to 
annually. Additionally, the report submission date would change from 60 
days after January 1 and July 1 of each year to before May 1 of each 
year. The changes to the reporting period and report date are being 
made to implement the August 31, 1992, amendment to 10 CFR 50.36a. The 
affected Technical Specifications Sections are 1.18, 3.11.1.4, 
3.11.2.6, 6.9.1.7, 6.14c, and the Index.
    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 a significant hazards 
consideration because operation of Callaway Plant with these changes 
would not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes do not affect accident initiators or 
assumptions. The radiological consequences of any accident 
previously evaluated remain unchanged.
    (2)Create the possibility of a new or different kind of accident 
from any previously evaluated.
    These changes do not impact any administrative controls nor do 
they involve physical alterations to the plant with respect to 
radioactive effluent. There is no new type of accident or 
malfunction created and the method and manner of plant operation 
will not change.
    (3) Involve a significant reduction in a margin of safety.
    The margin of safety remains unaffected since no design change 
is made and plant operation remains the same. The proposed changes 
do not affect any safety limits or boundary or system performance.
    As discussed above, the proposed changes are strictly 
administrative in nature and have no affect on plant operations. 
They do not involve a significant increase in the probability or 
consequences of an accident previously evaluated or create the 
possibility of a new or different kind of accident from any 
previously evaluated. These changes do not result in a significant 
reduction in a margin of safety. Therefore, it has been determined 
that the proposed changes do not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
    NRC Project Director: John N. Hannon

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

    Date of amendment request: October 6, 1993
    Description of amendment request: The proposed amendment would 
revise Technical Specifications Section 3.8.3, Electrical Power Systems 
- Onsite Power Distribution, to make the limiting conditions for 
operation for four emergency busses (NG05E, NG06E, NG07, and NG08) 
consistent with other technical specifications. The proposed revision 
would make the allowed outage time (AOT) for any of these emergency 
busses 72 hours. This is equivalent to the AOT for one train of the ESW 
per Technical Specification 3.7.4 and equivalent to the AOT for one 
train of the UHS cooling tower per Technical Specification 3.7.5.
    This amendment request also proposes an editorial change by 
removing the number sign () before each electrical bus, 
battery, and battery charger listed in Technical Specifications Section 
3.8.3 in order to clarify the specifications and make the nomenclature 
consistent with other sections.
    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 a significant hazards 
consideration because operation of the Callaway Plant with these 
changes would not:
    (1)Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The implementation of the proposed technical specification 
changes does not involve any modifications to the physical plant. 
Even though the MCCs themselves will have an allowed outage time of 
72 hours instead of 8 hours, the operability requirements of the ESW 
system itself have not been lessened. The addition of LCs NG07 and 
NG08 to the technical specifications and surveillances serves to 
clarify the 480-volt power supply requirements in the technical 
specifications. The proposed changes do not affect accident 
initiators or assumptions. The radiological consequences of any 
accident previously evaluated remain unchanged.
    (2)Create the possibility of a new or different kind of accident 
from any previously evaluated.
    As noted above, the proposed change eliminates inconsistent 
requirements from the technical specifications, but overall does not 
lessen the requirements on ESW system operability imposed by the 
technical specifications. The implementation of the proposed 
technical specification changes do not involve any modifications to 
the physical plant or any significant change to the methods of 
operation of plant systems. The proposed changes do not create any 
new accident initiators.
    (3)Involve a significant reduction in a margin of safety.
    The requirements of Technical Specification 3.7.4, Plant Systems 
- Essential Service Water System, provide specific limiting 
conditions for operation applicable to the ESW System. In accordance 
with the definition of operability contained in the technical 
specifications, the operability of the ESW MCCs has always been 
included within these requirements. The existing technical 
specification requirements for onsite A.C. power distribution 
systems are intended to assure the availability of A.C. power 
sources supplying multiple safety systems. The NG05E and NG06E MCCs 
identified by this proposed change provide power for a single safety 
system (ESW) and associated equipment. The use of the 72 hour limit 
for the ESW MCCs is consistent with the requirements of Regulatory 
Guide 1.93, ``Availability of Electrical Power Sources'' and has an 
insignificant impact on the Callaway Probabilistic Risk Analysis. 
LCs NG07 and NG08 also only provide power for a single safety system 
(ESW) and associated equipment (UHS cooling tower). Since the 
technical specification requirements relative to the ESW system 
operability are not lessened by this change, there will be no 
reduction in the margin of safety as defined in the basis for the 
technical specifications.
    As discussed, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated or create the possibility of a new or different 
kind of accident from any previously evaluated. These changes do not 
result in a significant reduction in a margin of safety. Therefore, 
it has been determined that the proposed changes do not involve a 
significant hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
    NRC Project Director: John N. Hannon

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

    Date of amendment request: February 1, 1994
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specifications 
(TS) by removing the review of the Emergency Plan and its implementing 
procedures from the list of responsibilities of the Plant Operations 
Review Committee (PORC). Guidance for this change was provided in 
Generic Letter 93-07, ``Modification of the Technical Specification 
Administrative Control Requirements for Emergency and Security Plans,'' 
dated December 28, 1993. Several other administrative TS changes are 
proposed including removing specific titles from the list of PORC 
members in TS 6.5.a.2 and deleting TS 6.5.b which describes the 
Corporate Support Staff.
    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 were revised in accordance with the 
provision of 10 CFR 50.92 to show no significant hazards exist. The 
proposed changes will not:
    1) involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The likelihood that an accident will occur is neither increased 
or decreased by these TS changes. These TS changes will not impact 
the function or method of operation of plant equipment. Thus, there 
is not a significant increase in the probability of a previously 
analyzed accident due to these changes. No systems, equipment, or 
components are affected by the proposed changes. Thus, the 
consequences of the malfunction of equipment important to safety 
previously evaluated in the Updated Safety Analysis Report (USAR) 
are not increased by these changes.
    The proposed changes are administrative in nature and, 
therefore, have no impact on accident initiators or plant equipment, 
and thus, do not affect the probabilities or consequences of an 
accident.
    2)create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Operation of the facility in accordance with the proposed TS 
changes would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed changes do not involve changes to the physical 
plant or operations. Since these administrative changes do not 
contribute to accident initiation, they do not produce a new 
accident scenario or produce a new type of equipment malfunction. 
Also, these changes do not alter any existing accident scenarios; 
they do not affect equipment or its operation, and thus, do not 
create the possibility of a new or different kind of accident.
    3)involve a significant reduction in the margin of safety.
    Operation of the facility in accordance with the proposed TS 
would not involve a significant reduction in a margin of safety. The 
proposed changes do not affect the plant equipment or operation. The 
requirements previously contained in the TS's that are being deleted 
are redundant and are contained in other controlled documents. 
Safety limits and limiting safety system settings are not affected 
by these proposed changes.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497.
    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.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis 
Obispo County, California

    Date of application for amendments: January 10, 1994, as 
supplemented February 3, 1994 (Reference LAR 94-01)
    Brief description of amendment request: The proposed amendments 
would revise the combined Technical Specifications (TS) for the Diablo 
Canyon Power Plant Unit Nos. 1 and 2 to change TS 3/4.3.2, ``Engineered 
Safety Features Actuation System Instrumentation,'' and TS 3/4.6.2.3, 
``Containment Cooling System.'' TS 3/4.3.2 would be revised to expand 
the mode applicability to include Mode 4 for the high-high containment 
pressure signal. TS 3/4.6.2.3 would be revised to clarify acceptable 
containment fan cooling unit (CFCU) configurations that satisfy the 
safety analysis requirements and to clarify the minimum required 
component cooling water flow supplied to the CFCU cooling coils.
    Date of individual notice in Federal Register: January 28, 1994 (59 
FR 4121)
    Expiration date of individual notice: February 28, 1994
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

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.

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 application for amendments: November 11, 1993
    Brief description of amendments: The amendments revise the 
Technical Specifications (TSs) for both Units 1 and 2 by relocating the 
tables of response time limits for the Reactor Protection System and 
the Engineered Safety Features Actuation System instruments from the 
TSs to the Updated Final Safety Analysis Report. These amendments are a 
``line-item'' TSs improvement and follow the guidance of Generic Letter 
93-08, ``Relocation of Technical Specification Tables of Instrument 
Response Time Limits.''
    Date of issuance: February 10, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.

    Amendment Nos.:  184 and 161
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67841) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated February 10, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick 
County, North Carolina

    Dates of application for amendments: December 31, 1992, as 
supplemented June 10, 1993, and August 23, 1993, and December 8, 1993.
    Brief description of amendments: The amendments change the 
Technical Specifications to (1) revise the definition of core 
alteration in section 1.0, Definitions, (2) clarify the TS 3/4.9.3, 
Control Rod Position, in the action statement, surveillance 
requirements and associated bases, and (3) revise the frequency for the 
channel calibration of the High Pressure Core Injection Steam Line 
Tunnel Temperature - High instrument.
    Date of issuance: February 8, 1994
    Effective date: February 8, 1994

    Amendment Nos.:  168 and 199
    Facility Operating License Nos. DPR-71 and DPR-62. Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: July 7, 1993 ( 56 FR 
36426), and January 5, 1994 (59 FR 617). The June 10, 1993, and August 
23, 1993, letters provided supplemental information and updated TS 
pages and did not change the initial proposed no significant hazards 
consideration determinations. The Commission's related evaluation of 
the amendments is contained in a Safety Evaluation dated February 8, 
1994.No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick 
County, North Carolina

    Date of application for amendments: January 4, 1991, as 
supplemented on June 24, 1991, December 19, 1991, and October 15, 1993.
    Brief description of amendments: The amendments (a) replace the 
current fire protection license condition in
    Facility Operating License Nos. DPR-71 and DPR-62 with the standard 
license conditon in Generic Letter 86-10 and (b) change the Technical 
Specifications to relocate the fire protection requirements to the 
BSEP, Units 1 and 2, Updated Final Safety Analysis Report.
    Date of issuance: February 10, 1994
    Effective date: February 10, 1994

    Amendment Nos.:  169 and 200
    Facility Operating License Nos. DPR-71 and DPR-62. The amendments 
replace the current fire protection license condition in
    Facility Operating License Nos. DPR-71 and DPR-62 with the standard 
license conditon in NRC Generic Letter 86-10, ``Implementation of Fire 
Protection Requirements.''
    Date of notices in Federal Register: March 20, 1991 (56 FR 11722) 
and February 5, 1992 (57 FR 4485) The Commission's related evaluation 
of the amendments is contained in a Safety Evaluation dated February 
10, 1994.No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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: July 26, 1993
    Brief description of amendment: The amendment makes three specific 
changes in the TS: (1) incorporates the auxiliary feedwater (AFW) flow 
control valve (FCV) automatic opening feature in periodic surveillance 
testing, and clarifies in the AFW Bases that given the FCVs auto-open 
design feature, (2) deletes periodic surveillance testing of the auto-
closure feature for the AFW motor-driven pump recirculation line 
valves; and (3) revises the general description of the AFW Bases so 
they are more concise and address directly the basis of the 
surveillance requirements.
    Date of issuance: February 14, 1994
    Effective date: February 14, 1994
    Amendment No.: 42
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 1, 1993 (58 
FR 46225) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 14, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.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. 50-295 and 50-304, Zion 
Nuclear Power Station Units 1 and 2, Lake County, Illinois

    Date of application for amendments: November 19, 1993
    Brief description of amendments: The amendment revises the 
Technical Specifications by changing the reactor vessel low temperature 
overpressure protection setpoint.
    Date of issuance: February 14, 1994
    Effective date: February 14, 1994
    Amendment Nos.: 153 and 141
    Facility Operating License Nos. DPR-39 and DPR-48. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67842) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 14, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

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

    Date of application for amendment: August 9, 1991, as supplemented 
by letters dated February 12, 1992, November 8, 1993, and January 25, 
1994.
    Brief description of amendment: The amendment would revise the 
Technical Specifications to delete the surveillance requirements and 
limiting operating conditions for the independent electrical turbine 
overspeed protection system and to extend the surveillance test 
interval for the turbine stop and control valves from monthly to an 
interval of not greater than yearly. Also included is a minor 
correction to a typographical error.
    Date of issuance: February 8, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 168
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 16, 1991 (56 FR 
51922) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 8, 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.

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 4, 1993
    Brief description of amendments: The amendments change the 
Technical Specifications to allow extended outage time for each train 
of the control area ventilation system to allow system maintenance to 
improve system reliability. The one time extension to 14 days (for each 
train, one at a time) will allow completion of the maintenance 
activities while one or both units are on-line; otherwise, it would be 
necessary to shut down both units to complete the maintenance 
activities or to divide the maintenance activities into less than 7-day 
segments, which would increase unavailability of the control area 
ventilation system.
    Date of issuance: February 10, 1994
    Effective date: February 10, 1994
    Amendment Nos.: 140 and 122
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62155) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 10, 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. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina 
Date of application of amendments: November 11, 1993, as 
supplemented November 22, 1993

    Brief description of amendments: The amendments provide an interim 
acceptance criteria for control rod drop time on Oconee, Unit 1.
    Date of Issuance: February 9, 1994
    Effective date: February 9, 1994
    Amendment Nos.: 205, 205, and 202
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: November 29, 1993 (58 
FR 62689) The November 22, 1993, letter provided clarifying information 
that did not change the scope of the November 11, 1993, application and 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 9, 1994. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, 
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, 
Pennsylvania

    Date of application for amendments: June 14, 1990, as supplemented 
November 17, 1993
    Brief description of amendments: These amendments revise the 
Electrical Power System Shutdown, the AC Distribution - Shutdown, and 
the DC Distribution - Shutdown Specifications to more closely resemble 
the wording contained in the Standard Technical Specifications. The 
November 17, 1993, supplement changed existing terminology used to 
designate two emergency busses in Unit No. 1 and two DC busses in Unit 
2 to standard nomenclature.
    Date of issuance: February 7, 1994
    Effective date: February 7, 1994
    Amendment Nos.:  180 and 60
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 19, 1990 (55 
FR 38601) The November 17, 1993, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated February 7, 
1994.No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

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

    Date of application for amendment: August 23, 1993
    Brief description of amendment: This amendment will delete the 
option of using a movable incore detector to determine Incore 
Instrumentation System operability from the provisions of Technical 
Specification 3.3.3.2.

    Date of issuance: February 8, 1994
    Effective date: February 8, 1994
    Amendment No.: 64
    Facility Operating License No. DPR-67: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 13, 1993 (58 FR 
52985) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 8, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003

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

    Date of applications for amendment: May 26 and December 2, 1993
    Brief description of amendment: The amendment revises the TMI-1 
Technical Specifications to correct the definition of flood stage. The 
amendment also revises the TMI-1 Technical Specifications to remove the 
limiting conditions for operation and surveillance requirements for the 
Chlorine Detection Systems. Because this bridge was underwater during 
the 1972 flooding, the reference datum point location will be specified 
as the Susquehanna River Gage at Harrisburg. TMI-1 removed the gaseous 
chlorine system for the Circulating Water and River Water Systems.
    Date of issuance: February 10, 1994
    Effective date: As of its date of issuance.
    Amendment No.: 182
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59750) and January 5, 1994 (59 FR 621).The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
February 10, 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.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of application for amendments: December 16, 1992, as 
supplemented December 22, 1993.
    Brief description of amendments: The amendments revise the licenses 
to allow the replacement of portions of the current Reactor Protection 
System instrumentation with a digital signal processing system.
    Date of issuance: February 7, 1994
    Effective date: February 7, 1994
    Amendment Nos.: 175 & 160
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments add a 
license condition to the Operating Licenses.
    Date of initial notice in Federal Register: March 3, 1993 (58 FR 
12263) The December 22, 1993, letter provided clarifying information 
which did not change the staff's initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
February 7, 1994. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

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

    Date of application for amendment: August 4, 1993
    Brief description of amendment: The amendment incorporates an 
additional Emergency Diesel Generator Surveillance Requirement, 
4.8.1.1.2.C.8, items a, b, and c, to the Technical Specification 
Section 3/4.8, ``Electrical Power Systems.'' The change requires 
starting the EDG, with offsite power available, as a result of a Safety 
Injection Actuation Signal.
    Date of issuance: February 14, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 171
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67852) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 14, 1994. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Thames Valley State Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360.

Pennsylvania Power and Light Company, Docket No. 50-388, 
Susquehanna Steam Electric Station, Unit 2, Luzerne County, 
Pennsylvania

    Date of application for amendment: August 19, 1992, as supplemented 
by letters dated May 18 and October 7, 1993
    Brief description of amendment: The amendment changed the Technical 
Specifications to revise the logic which controls the automatic 
transfer of the High Pressure Coolant Injection pump suction source on 
high suppression pool level.
    Date of issuance: February 9, 1994
    Effective date: February 9, 1994
    Amendment No.: 101
    Facility Operating License No. NPF-22. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 16, 1992 (57 
FR 42778) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 9, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

Philadelphia Electric Company, Docket No. 50-352, Limerick 
Generating Station, Unit 1, Montgomery County, Pennsylvania

    Date of application for amendment: August 27, 1993, supplemented by 
letter dated November 17, 1993
    Brief description of amendment: The amendment allows an expanded 
operating domain for the Limerick Generating Station (LGS), Unit 1, 
resulting from the implementation of the Average Power Range Monitor - 
Rod Block Monitor Technical Specifications/Maximum Extended Load Line 
Limit Analysis.
    Date of issuance: February 10, 1994
    Effective date: February 10, 1994
    Amendment No. 66
    Facility Operating License No. NPF-39. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 13, 1993 (58 FR 
52992) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 10, 1994. No 
significant hazards consideration comments received:
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Docket No. 50-352, Limerick 
Generating Station, Unit 1, Montgomery County, Pennsylvania

    Date of application for amendment: November 30, 1993
    Brief description of amendment: This amendment changes the Appendix 
A technical specifications by allowing the third Type A Containment 
Integrated Leakage Rate Test in the first 10-year service period to be 
conducted at Refuel 6.
    Date of issuance: February 16, 1994
    Effective date: February 16, 1994
    Amendment No. 67
    Facility Operating License No. NPF-39. The amendment revised the 
Technical Specification.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67858) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 16, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, 
Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania

    Date of application for amendments: November 30, 1993
    Brief description of amendments: These amendments decrease the test 
frequency of the drywell-to-suppression chamber bypass leak test to 
coincide with the primary Containment Integrated Leak Rate Test 
interval and require an additional test to measure the vacuum breaker 
leakage area for those outages for which the drywell-to-suppression 
chamber bypass test is not scheduled.
    Date of issuance: February 17, 1994
    Effective date: February 17, 1994
    Amendment Nos. 68 and 31
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 5, 1994 (59 FR 
626) The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 17, 1994.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendment: September 28, 1992
    Brief description of amendment: The amendment revises the flow 
requirement for the Core Spray (CS) pumps and the associated Bases. The 
change reduces the CS pump minimum flow acceptance criteria by 10% and 
addresses an inconsistency between the system leakage rates in the 
Updated Final Safety Analysis Report and the Technical Specifications 
(TS). Specifically, the surveillance testing required by the TS is 
intended to verify the capability of a core spray pump to deliver 
acceptable flow to the core. The new CS pump minimum flow acceptance 
criteria now accounts for system leakage that is not delivered to the 
core.
    Date of issuance: February 8, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 204
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 9, 1992 (57 FR 
58250) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 8, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Southern California Edison Company, et al., Docket Nos. 50-361 and 
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, 
San Diego County, California

    Date of application for amendments: April 7, 1992
    Brief description of amendments: These amendments revise Technical 
Specifications Tables 3.3-3, 3.3-4, 3.3-5, and 4.3-2, which provide the 
requirements for the Engineered Safety Features Actuation System 
(ESFAS) instrumentation. This Technical Specification change will 
clarify that a Manual Safety Injection Actuation Signal does not 
actuate a Containment Cooling Actuation Signal. This is an editorial 
change to make the Technical Specifications consistent with plant 
design.
    Date of issuance: February 4, 1994
    Effective date: February 4, 1994
    Amendment Nos.: 110 and 99
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 10, 1992 (57 FR 
24679) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 4, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713

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: March 10, 1993; amended January 
31, 1994 (TS 93-02)
    Brief description of amendments: The amendments add a reference to 
the test requirements of 10 CFR 50, Appendix J, ``Primary Reactor 
Containment Leakage Testing for Water-Cooled Power Reactors'' to the 
technical specifications at various locations, and remove the 
corresponding detailed test requirements and acceptance criteria. Other 
containment system specifications related to this issue are also 
removed. In addition, a change to Table 3.6-2, ``Containment Isolation 
Valves,'' clarifies the additional testing requirements for the 
containment purge valves.
    Date of issuance: February 10, 1994
    Effective date: February 10, 1994
    Amendment Nos.: 176, Unit 1 - 167, Unit 2
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: May 12, 1993 (58 FR 
28059) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 10, 1994.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of application for amendment: September 23, 1991
    Brief description of amendment: This amendment allows an alternate 
method for verifying whether a control rod drive pump is operating.
    Date of issuance: February 14, 1994
    Effective date: February 14, 1994
    Amendment No. 55
    Facility Operating License No. NPF-58. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 13, 1991 (56 
FR 57705) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 14, 1994. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.

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: December 10, 1993
    Brief description of amendments: The amendments modify the 
surveillance frequency of the Auxiliary Feedwater System pumps from 
monthly to quarterly.
    Date of issuance: February 7, 1994
    Effective date: February 7, 1994
    Amendment Nos.:  177 and 158
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: January 5, 1994 (59 FR 
631) The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 7, 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.

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

    Date of application for amendment: February 23, 1993
    Brief description of amendment: The amendment revises TS Section 
3.5, ``Instrumentation System,'' Table TS 3.5-6, ``Instrumentation 
Operating Conditions for Indication,'' and Table TS 4.4-1, ``Minimum 
Frequencies for Checks, Calibrations and Test of Instrument Channels.'' 
The amendment adds operability and surveillance requirements for the 
reactor vessel level indication and core exit thermocouple 
instrumentation to satisfy the recommendations of Generic Letter 83-37, 
``NUREG-0737 Technical Specifications.'' Similar additions are made for 
the wide range steam generator level instrumentation to satisfy 
Regulatory Guide 1.97 recommendations. Administrative changes are also 
incorporated as part of converting the TS document to the WordPerfect 
software.
    Date of issuance: February 9, 1994
    Effective date: February 9, 1994
    Amendment No.: 105
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 21, 1993 (58 FR 
39061) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 9, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.

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

    Date of application for amendment: November 16, 1993, as 
supplemented on December 7, 1993.
    Brief description of amendment: The amendment modifies KNPP TS 
4.4.a.7 by deleting the requirement that couples the performance of the 
Type A leakage tests to the 10-year inservice inspection program 
requirements. This change was made to reflect the partial exemption 
from the requirements of 10 CFR 50, Appendix J, Section III.D.a.(a), 
which was granted by the NRC on February 14, 1994. In addition, 
administrative changes to KNPP TS Section 4.4 and its associated bases 
have been made.
    Date of issuance: February 17, 1994
    Effective date: February 17, 1994
    Amendment No.: 106
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67865) The December 7, 1993, submittal provided additional 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated February 17, 1994.No significant hazards consideration comments 
received: No.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301 
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks 
Manitowoc County, Wisconsin

    Date of application for amendments: March 24, 1993
    Brief description of amendments: These amendments revised Technical 
Specifications (TS) Section 15.6 to update several position titles, to 
modify the composition and duties of the Manager's Supervisory Staff 
(the onsite review committee), and to remove a redundant review of the 
Facility Fire Protection Program implementing procedures.
    Date of issuance: January 27, 1994
    Effective date: January 27, 1994
    Amendment Nos.: 146 and 150
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 18, 1993 (58 FR 
43940) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 27, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas

    Date of amendment request: May 27, 1993
    Brief description of amendment: The proposed changes would revise 
the heatup, cooldown, and cold overpressure mitigation system power-
operated relief valve setpoint pressure/temperature limits. The revised 
limits reflect the analysis of the most recently withdrawn surveillance 
capsule associated with the reactor vessel radiation surveillance 
program (10 CFR 50, Appendix H). The revised limits bound operation 
through 13.6 Effective Full Power Years (EFPY).
    Date of issuance: February 10, 1994
    Effective date: February 10, 1994, to be implemented within 30 days 
of issuance.
    Amendment No.: 71
    Facility Operating License No. NPF-42: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36449) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 10, 1994.No significant 
hazards consideration comments received: No.Local Public Document Room 
Locations: Emporia State University, William Allen White Library, 1200 
Commercial Street, Emporia, Kansas 66801 and Washburn University School 
of Law Library, Topeka, Kansas 66621
    Dated at Rockville, Maryland, this 23rd day of February 1994.
    For the Nuclear Regulatory Commission
Jack W. Roe,
Director, Divisio Director, Division of Reactor Projects - III/IV/V, 
Office of Nuclear Reactor Regulation
[Doc. 94-4562 Filed 3-1-94; 8:45 am]
BILLING CODE 7590-01-F