[Federal Register Volume 62, Number 204 (Wednesday, October 22, 1997)]
[Notices]
[Pages 54866-54885]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-11022]


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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

[[Page 54867]]

issued, under a new provision of section 189 of the Act. This provision 
grants the Commission the authority to issue and make immediately 
effective any amendment to an operating license upon a determination by 
the Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from September 29, 1997, through October 9, 1997. 
The last biweekly notice was published on October 8, 1997 (62 FR 
52578).

Notice Of Consideration Of Issuance Of Amendments To Facility 
Operating Licenses, Proposed No Significant Hazards Consideration 
Determination, And Opportunity For A Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By November 21, 1997, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for 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

[[Page 54868]]

Commission may issue the amendment and make it immediately effective, 
notwithstanding the request for a hearing. Any hearing held would take 
place after issuance of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendment request: April 7, 1997, as supplemented on August 
7, 1997.
    Description of amendment request: The proposed amendment would 
revise the plants' technical specifications to permit replacement of 
the 125 volt dc Gould batteries with new C&D Charter Power Systems, 
Inc., batteries.
    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 change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The replacement C&D battery has been selected to meet or exceed 
the design, functional, and operational requirements of those of the 
present Gould battery, including crosstie load limitations. The C&D 
batteries are similar in design to the installed Gould batteries 
(e.g., electrolyte specific gravity and construction of the plates) 
except for capacity. The replacement C&D batteries have a 
significantly larger capacity than the Gould batteries, which can 
provide additional margin for future use. Also, the C&D batteries 
are qualified for a 20 year life and meet the latest applicable 
standards. The short circuit current provided by the C&D batteries 
is well within the interrupting capability of the existing DC system 
circuit breakers.
    Additionally, the crosstie limit is increased to take advantage 
of the larger C&D battery capacity. The C&D batteries were sized 
based on having sufficient capacity to energize the design basis DC 
loads for an operating unit with the IEEE-485 design margin while 
maintaining the desired limited DC load of 200 amps for a shutdown 
unit. This proposed change allows use of the C&D batteries' larger 
capacity. The overall design, function, and operation of the DC 
system and equipment has not been altered by these changes. The 
proposed changes do not affect any accident initiators or precursors 
and do not alter the design assumptions for the systems or 
components used to mitigate the consequences of an accident as 
analyzed in UFSAR Chapter 15. Therefore, there is no increase in the 
probability or consequences of an accident previously evaluated.
    B. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The replacement C&D batteries will provide the same functions as 
those of the installed Gould batteries and will be operated with the 
same types of operational controls. These limits include battery 
float terminal voltage, individual cell voltage and electrolyte 
specific gravity, and crosstie loading. Crosstie conditions are 
allowed under the present Technical Specifications. The crosstie 
limit is increased to take advantage of the larger C&D battery 
capacity. The remaining changes are administrative in nature or 
provide clarification to maintain consistency with other Technical 
Specifications.
    The DC system and its equipment will continue to perform the 
same functions and be operated in the same fashion. The proposed 
change does not create any new or common failure modes. The proposed 
changes do not introduce any new accident initiators or precursors, 
or any new design assumptions for the systems or components used to 
mitigate the consequences of an accident. Therefore, the possibility 
of a new or different kind of accident from any accident previously 
evaluated has not been created.
    C. The proposed change does not involve a significant reduction 
in a margin of safety.
    The replacement C&D batteries will meet or exceed the design, 
functional, and qualification requirements [of] those of the 
installed Gould batteries. The proposed Technical Specification 
limitations for the C&D batteries are derived from the same 
methodology as the Gould batteries with applied margins in 
accordance with IEEE-485. Increasing the crosstie loading limit 
takes advantage of the larger C&D battery capacity with its 
increased design margin. The proposed change to the crosstie loading 
limit will continue to conservatively envelope the postulated design 
requirements. The remaining changes are administrative in nature or 
provide clarification to maintain consistency with other Technical 
Specifications.
    The inherent design conservatism of the DC system and its 
equipment has not been altered. The DC system and its equipment will 
continue to be operated with the same degree of conservatism. 
Therefore, there is no reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Byron Public Library District, 
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois

    Date of amendment request: June 30, 1997, as supplemented on 
September 25, 1997.
    Description of amendment request: The proposed amendment would 
revise the plants' technical specifications to permit the licensee to 
take credit for soluble boron in spent fuel storage pool water to 
maintain an acceptable margin of subcriticality.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The following accidents have been specifically evaluated 
relative to the SFP [spent fuel pool]: fuel assembly drop, 
accidental misloading of spent fuel

[[Page 54869]]

assemblies into the SFP racks, and loss of normal cooling.
    There is no increase in the probability of a fuel assembly drop 
accident in the SFP when considering the presence of soluble boron 
in the SFP water for criticality control. The handling of the fuel 
assemblies in the SFP has previously been performed in borated 
water. The criticality analysis shows the consequences of a fuel 
assembly drop accident in the SFP are not affected when considering 
the presence of soluble boron.
    There is no increase in the probability of the accidental 
misloading of spent fuel assemblies into the SFP racks when 
considering the presence of soluble boron in the pool water for 
criticality control. Fuel assembly placement will continue to be 
controlled in accordance with approved fuel handling procedures and 
the spent fuel storage configuration limitations. Periodic 
surveillances of the SFP inventory (physical inventory and piece 
counts) are performed in accordance with station procedures. These 
surveillances ensure physical SFP inventory verification is 
performed at least once per year and in a timely manner upon 
completion of fuel movement in the SFP. The addition of credit for 
decay time in the spent fuel pool in determining allowable storage 
requirements is an extension of the reactivity equivalencing 
methodologies used for burnup credit in WCAP-14416-NP-A, 
``Westinghouse Spent Fuel Rack Criticality Analysis Methodology,'' 
Revision 1, November 1996.
    There is no increase in the consequences of the accidental 
misloading of spent fuel assemblies into the SFP racks because 
criticality analyses demonstrate that the pool will remain 
subcritical following an accidental misloading if the pool contains 
an adequate boron concentration. The proposed TS limitations and 
surveillance frequency will ensure that an adequate SFP boron 
concentration is maintained.
    There is no increase in the probability of the loss of normal 
cooling to the SFP water when considering the presence of soluble 
boron in the pool water for subcriticality control since a high 
concentration of soluble boron has previously been maintained in the 
SFP water. A loss of normal cooling to the SFP water causes an 
increase in the temperature of the water passing through the stored 
fuel assemblies. This causes a decrease in water density which would 
result in a decrease in reactivity when Boraflex neutron absorber 
panels are present in the racks. However, since the proposed change 
does not consider Boraflex to be present in the racks, and the SFP 
water has a high concentration of boron, a density decrease causes a 
positive reactivity addition. [The] consequences of this accident 
are bounded by the misloaded assembly analysis. Because adequate 
soluble boron will be maintained in the SFP water, the consequences 
of a loss of normal cooling to the SFP will not be increased.
    The proposed 48 hour surveillance frequency will be used to 
verify the boron concentration is within the initial assumptions of 
the criticality analysis. The current frequency of 24 hours was 
based on the sampling frequency for reactor coolant system (RCS) 
shutdown margin in Mode 5. A dilution of the SFP to a 
k<INF>eff</INF> greater than 0.95 would take a much longer time than 
an RCS dilution resulting in loss of shutdown margin. This is due to 
the larger SFP volume compared to the RCS volume, and the turnover 
rate of water in the SFP is much less due to the lack of large 
dilution sources for the SFP. The 48 hour sampling frequency is 
sufficient based on operating experience, and based on the fact that 
significant changes in the boron concentration in the spent SFP are 
difficult to produce without detection, due to the large inventory 
of water. Soluble boron concentration reduction requires the inflow 
and outflow of large volumes of water which are readily detected by 
SFP and fuel handling building sump high level alarms, flooding in 
the fuel handling building or by normal operator rounds through the 
SFP area (once every eight hours), allowing adequate time for 
operator intervention prior to exceeding a k<INF>eff</INF> of 0.95. 
Therefore, consequences of an accident previously evaluated are not 
increased by the change in surveillance frequency.
    The format revisions to Specification 5.6.1.1 and reference to 
the report containing the specific NRC-approved criticality 
methodology in Specification 6.9.1.10 are administrative in nature 
and will not result in an increase in the probability or 
consequences of an accident previously evaluated.
    Therefore, based on the above analysis, the proposed changes 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The results of criticality accident analyses in the SFP are 
discussed in the UFSAR [Updated Final Safety Analysis Report] and in 
Criticality Analysis Reports associated with previous licensing 
activities. Specific accidents considered include fuel assembly 
drop, accidental misloading of spent fuel assemblies into the SFP 
racks, and loss of normal cooling.
    LCO 3.9.1, ``BORON CONCENTRATION,'' contains limitations on the 
boron concentration in the filled portions of the reactor coolant 
system and the refueling canal during Mode 6. ComEd has maintained 
soluble boron in the SFP at all times and has imposed administrative 
limits on the SFP boron concentration, due in part to this 
requirement. LCO 3.9.11 establishes specific boron concentration 
requirements for the SFP water consistent with the results of the 
new criticality analysis based on the NRC-approved methodology of 
WCAP-14416-NP-A, ``Westinghouse Spent Fuel Rack Criticality Analysis 
Methodology,'' Revision 1, November 1996. Credit is also taken for 
radioactive decay time of the spent fuel.
    Since soluble boron has always been maintained in the SFP water 
and is currently controlled administratively, the implementation of 
this requirement will have little effect on normal pool operations 
and maintenance. The implementation of the proposed limitations on 
the SFP boron concentration will only result in a requirement to 
verify boron concentration of the SFP water every 48 hours rather 
than every 24 hours. Sampling every 48 hours is sufficient to verify 
the SFP boron concentration meets the assumptions of the criticality 
analysis.
    Because soluble boron has always been present in the SFP and has 
been administratively controlled, a dilution of the SFP soluble 
boron has always been a possibility. As shown in the SFP dilution 
evaluation performed for Byron and Braidwood, a dilution of the SFP 
which could increase the rack k<INF>eff</INF> to greater than 0.95 
(i.e., which could reduce the required margin to criticality) is not 
a credible event.
    Therefore, the implementation of the proposed limitations on the 
SFP boron concentration and surveillance frequency will not result 
in the possibility of a new kind of accident.
    The proposed change to Specification 5.6.1.1 identifies the 
requirements for the spent fuel rack storage configurations. The 
proposed changes relate to the criteria for determining the storage 
configuration. Since the proposed SFP storage configuration 
limitations will be similar to those currently in the Byron and 
Braidwood TS, these limitations will not have any significant effect 
on normal SFP operations and maintenance and will not create any 
possibility of a new or different kind of accident. Verifications 
will continue to be performed to ensure that the SFP loading 
configuration meets specified requirements.
    The format revisions to Specification 5.6.1.1 and reference to 
the report containing the specific NRC-approved criticality 
methodology in Specification 6.9.1.10 are administrative in nature 
and will not create the possibility of a new [or] different kind of 
accident.
    As discussed above, there is no significant change in plant 
configuration or equipment and the proposed changes will not create 
the possibility of a new or different kind of accident.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed TS changes and the resulting spent fuel storage 
operating limits will provide adequate safety margin to ensure that 
the stored fuel assembly array will always remain subcritical. These 
limits are based on a plant specific criticality analysis performed 
in accordance with the NRC-approved Westinghouse spent fuel rack 
criticality analysis methodology (WCAP-14416-NP-A). Credit is also 
taken for radioactive decay time of the spent fuel.
    Soluble boron credit provides significant negative reactivity in 
the SFP such that the k<INF>eff</INF> is maintained less than or 
equal to 0.95. The proposed surveillance frequency will be used to 
verify the boron concentration is within the initial assumptions of 
the criticality analysis. A storage configuration has also been 
defined, with a 95-percent probability at a 95-percent confidence 
level, that ensures the spent fuel rack k<INF>eff</INF> will be less 
than 1.0 with no credit for soluble boron or Boraflex panels in the 
racks. In addition to soluble boron credit, credit is taken for fuel 
assembly burnup, decay time, and IFBAs [Integral Fuel Burnable 
Absorber] when determining assembly storage requirements.

[[Page 54870]]

    The loss of substantial amounts of soluble boron from the SFP 
which could lead to exceeding a k<INF>eff</INF> of 0.95 has been 
evaluated and shown not to be credible. These evaluations show that 
the dilution of the SFP boron concentration from 2000 ppm to 550 ppm 
is not credible and that the spent fuel rack k<INF>eff</INF> will 
remain less than 1.0 when flooded with unborated water.
    The format revisions to Specification 5.6.1.1 and reference to 
the report containing the specific NRC-approved criticality 
methodology in Specification 6.9.1.10 are administrative in nature 
and will not result in a significant reduction in the plant's margin 
of safety.
    Therefore, the proposed changes in this license amendment will 
not result in a significant reduction in the plant's 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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: August 12, 1997
    Description of amendment request: The proposed amendments would 
remove a Technical Specification surveillance requirement to verify 
that sediment deposition within the lake screenhouse is not greater 
than one foot in thickness. Control of sediment accumulation in the 
lake screenhouse would be accomplished through the Service Water 
Performance Monitoring Program.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1) Involve a significant increase in the probability or 
consequences of an accident previously identified because:
    Surveillance's [sic] to fully verify [that] the Ultimate Heat 
Sink contains enough water to perform its design function will 
continue. All cleanliness issues associated with ensuring 
operability of Core Standby Cooling System - Equipment Cooling Water 
System (CSCS-ECWS) equipment will be performed under the Service 
Water Performance Monitoring Program, which meets GL 89-13 
[<gr-thn-eq>Service Water System Problems Affecting Safety-Related 
Equipment<gr-thn-eq>] recommended actions. By performing these 
inspections per GL 89-13, LaSalle will ensure that there is no build 
up of sediment, which could hinder or impede the design operation of 
any safety or non-safety related equipment which takes a suction 
from the service water tunnel. Based on the nature of sediment, 
where it collects, and system design, the CSCS-ECWS will be 
available if called upon or started to respond in case of an 
accident for equipment cooling and long term cooling.
    At no time, during approximately fourteen years of LaSalle 
operation, has sediment built up or accumulated either in front of 
the inlet to the CSCS cooling water screen bypass supply line or the 
six 36-inch normal tunnel supply lines in such a manner that the 
flow of water through these lines could have been reduced or 
blocked. Instead, loose sediment collects in quiescent areas near 
the traveling screens, the north end of the Service Water Tunnel, 
under the outlets of the 36-inch normal tunnel supply lines in the 
service water tunnel, and downstream of the butterfly isolation 
valve in the 54 inch CSCS cooling water screen bypass supply line. 
The sediment that collects in the service water tunnel does not 
build up in a manner such that CSCS-ECWS, non-essential station 
service water, or fire pump suctions from the tunnel are affected, 
based on inspections since 1992.
    The CSCS equipment cooling bypass valve, OE12-F300, is the 
manual butterfly valve in the CSCS cooling water screen bypass 
supply line. The bypass valve is being added to the ASME Section XI 
Inservice Testing Program to cycle the valve quarterly. This valve 
cycling will help maintain sediment level in the bypass line at a 
low level due to flow through the line while the valve is not fully 
closed and thus assure the bypass line remains available. The flow 
is created due to the differential pressure across the circulating 
water traveling screens with circulating water pumps in operation.
    Therefore, neither essential nor non-essential service water 
will be lost due to sediment. Neither the probability nor the 
consequences of an accident are increased by the deletion of SR 
4.7.1.3.c.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    Inspections for sedimentation will continue to be required by 
LaSalle's Service Water System Performance Monitoring Program per GL 
89-13, to ensure continued operability of Core Standby Cooling 
System-Equipment Cooling Water System (CSCS-ECWS). The Ultimate Heat 
Sink operability requires assurance of a specific volume of water to 
provide cooling for at least 30 days for long term cooling following 
an accident. The public will be protected by the safety analysis in 
place by the fact that the safety and non-safety related equipment 
which take a suction from the service water tunnel will not be 
impaired by sediment. Therefore, there will be no possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3) Involve a significant reduction in the margin of safety 
because:
    The Ultimate Heat Sink continues to be demonstrated Operable by 
verifying a sufficient volume of water per TS SR 4.7.1.3.a and 
4.7.1.3.b. Equipment operability will still be required per 
Technical Specifications 3/4.7.1.1 and 3/4.7.1.2 for the CSCS-ECWS 
systems. Sedimentation in the lake screenhouse is a maintenance/
cleanliness issue addressed by the LaSalle Service Water Performance 
Monitoring Program. The program ensures equipment operability by 
both inspection for and removal of sedimentation and chemical 
control with a biocide to limit the growth of biological material 
and silt dispersant to help keep silt in the flow stream from 
coagulating. Therefore, there is minimal or no reduction in the 
margin of safety due to the deletion of this surveillance 
requirement.
    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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: September 29, 1997 (NRC-97-0089)
    Description of amendment request: The proposed amendment would 
relocate the requirements for selected instrumentation and the 
associated Bases from the technical specifications (TS) to the updated 
final safety analysis report. The affected instrumentation is seismic 
monitoring (TS 3.7.2), meteorological monitoring (TS 3.7.3), the 
traversing in-core probe system (TS 3.7.7), the chlorine detection 
system (TS 3.7.8), and the loose parts detection system (TS 3.7.10). 
Changes to the TS index and list of tables were also requested to 
reflect the relocation of these TS and associated Bases. NRC Generic 
Letter 95-10, ``Relocation of Selected Technical Specification 
Requirements Related to Instrumentation,'' dated December 15, 1995, 
provided information concerning relocation of the requirements for 
these instruments.

[[Page 54871]]

    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 changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes would relocate TS 3/4.3.7.2 - Seismic 
Monitoring Instrumentation, TS 3/4. 3.7.3 - Meteorological 
Monitoring Instrumentation, TS 3/4.3.7.7 - Traversing In-Core Probe 
System, TS 3/4.3.7.8 - Chlorine Detection System, and TS 3/4.3.7.10 
- Loose-Part Detection System and their associated Bases to the 
Fermi 2 Updated Final Safety Analysis Report (UFSAR). They would 
also delete the special reporting requirements from the 
aforementioned TS which contain such requirements. The proposed 
changes would revise the TS Index and List of Tables to reflect the 
relocation of these TS and associated Bases. The relocated TS 
changes would be controlled in accordance with the requirements of 
10 CFR 50.59.
    The proposed changes affect TS that do not meet the NRC's 
``Final Policy Statement on Technical Specification Improvements for 
Nuclear Power Reactors'' or 10 CFR 50.36(c)(2)(ii) criteria for 
inclusion in TS. These TS relocations are consistent with NUREG-
1433, ``Standard Technical Specifications, General Electric Plants, 
BWR/4,'' Revision 1, April 1995. Furthermore, these five TS are 
specifically identified in NRC Generic Letter 95-10, ``Relocation of 
Selected Technical Specifications Requirements Related to 
Instrumentation,'' dated December 15, 1995, as suitable for 
relocation to licensee-controlled documents.
    The Special Report requirements of TS 3/4.3.7.2, TS 3/4.3.7.3, 
and TS 3/4.3.7.10 would be deleted as part of their relocation to 
the UFSAR. The NRC reporting criteria of 10 CFR 50.72, ``Immediate 
Notification Requirements for Operating Nuclear Reactors,'' and 10 
CFR 50.73, ``Licensee Event Report Systems'' provide appropriate 
requirements for reporting degraded and non-conforming conditions to 
the NRC.
    These proposed TS changes do not involve a significant increase 
in the probability of an accident previously evaluated because no 
changes are being made to any accident initiator. No previously 
analyzed accident scenario is changed, and initiating conditions and 
assumptions remain as previously analyzed.
    These proposed TS changes do not involve a significant increase 
in the consequences of an accident previously evaluated because the 
proposed changes do not affect accident sequences or assumptions 
used in evaluating the radiological consequences of an accident. The 
proposed changes do not alter the source term, containment isolation 
or allowable radiological releases.
    2. The changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not change the way in which the plant is 
operated and no new or different failure modes have been defined for 
any plant system or component. No limiting single failure has been 
identified as a result of the proposed changes. No new or different 
types of failures or accident initiators are introduced by the 
proposed changes.
    3. The changes do not involve a significant reduction in the 
margin of safety.
    The proposed changes involve instrumentation and systems which 
are not inputs in the calculation of any safety margin with regard 
to Technical Specification Safety Limits, Limiting Safety System 
Settings, Limiting Control Settings or Limiting Conditions for 
Operation, or other previously defined margins for any structure, 
system, or component.
    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: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226
    NRC Project Director: John N. Hannon

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 amendment request: September 11, 1997
    Description of amendment request: The proposed amendments would 
relocate the reactor trip system and engineered safety feature 
actuation system response times from technical specification (TS) 
tables 3.3-2 and 3.3-5 to Section 3 of the licensee's Licensing 
Requirements Manual (LRM) in accordance with the guidance provided in 
NRC Generic Letter 93-08. Subsequent changes to the LRM would be 
controlled in accordance with the requirements of 10 CFR 50.59. The 
proposed amendments would also make several editorial changes in TSs 
3.3.1.1 and 3.3.1.2, as well as making conforming changes to the Bases 
for these TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed amendment relocates the instrument response time 
limits for the reactor trip system (RTS) and engineered safety 
feature actuation system (ESFAS) from the technical specifications 
to the Licensing Requirements Manual (LRM). The Core Operating 
Limits Report (COLR) and containment penetrations table (containment 
isolation valves) are controlled and maintained in the LRM. The LRM 
was developed to control and maintain those items removed from the 
technical specifications. The proposed amendment conforms to the 
guidance given in Enclosures 1 and 2 of Generic Letter 93-08. 
Neither the response time limits nor the surveillance requirements 
for performing response time testing will be altered by this 
submittal. The overall RTS and ESFAS functional capabilities will 
not be changed and assurance that action requirements of the 
protective and engineered safety features systems are completed 
within the time limits assumed in the accident analyses is 
unaffected by the proposed amendment. Therefore, operation of the 
facility in accordance with the proposed amendment will not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed amendment will not change the physical plant or the 
modes of plant operation defined in the operating license. The 
change does not involve the addition or modification of equipment 
nor does it alter the design or operation of plant systems. 
Therefore, operation of the facility in accordance with the proposed 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The measurement of instrumentation response times at the 
frequencies specified in the technical specification provides 
assurance that actions associated with the protective and engineered 
safety features systems are accomplished within the time limits 
assumed in the accident analyses. The response time limits, and the 
measurement frequencies remain unchanged by the proposed amendment. 
The proposed changes do not alter the basis for any other technical 
specification that is related to the establishment of or maintenance 
of a nuclear safety margin. Therefore, operation of the facility in 
accordance with 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: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts &

[[Page 54872]]

Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf 
Nuclear Station, Unit 1, Claiborne County, Mississippi

    Date of amendment request: September 18, 1997
    Description of amendment request: The amendment would decrease the 
safety limit for the minimum critical power ratio (MCPR) from 1.12 to 
1.11 for two recirculation loop operation and from 1.14 to 1.12 for 
single recirculation loop operation in Technical Specification (TS) 
2.1.1.2. Because the proposed amendment is for Cycle 10 operation, the 
amendment would also revise the footnotes to TSs 2.1.1.2 and 5.6.5 to 
state that the MCPR values and the items 19 and 20 are ``applicable 
only for Cycle 10 operation.'' Cycle 10 operation is after the next 
(i.e., 9th) refueling outage.
    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:
    I. The proposed change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The Minimum Critical Power Ratio (MCPR) safety limit is defined 
in the Bases to Technical Specification [TS] 2.1.1 as that limit 
which ``ensures that during normal operation and during Anticipated 
Operational Occurrences (AOOs), at least 99.9% of the fuel rods in 
the core do not experience transition boiling.'' The MCPR safety 
limit is re-evaluated for each reload and, for GGNS [Grand Gulf 
Nuclear Station, Unit 1] Cycle 10, the analyses have concluded that 
a two-loop MCPR safety limit of 1.11 based on the application of 
GE's [General Electric Company's] cycle-specific MCPR safety limit 
methodology is necessary to ensure that this acceptance criterion is 
satisfied. For single-loop operation, a MCPR safety limit of 1.12 
based on GE's cycle-specific MCPR safety limit methodology was 
determined to be necessary. Core MCPR operating limits are developed 
to support the Technical Specification [TS] 3.2 requirements and 
ensure these safety limits are maintained in the event of the worst 
case transient. Since the MCPR safety limit will be maintained at 
all times, operation under the proposed changes will ensure [that] 
at least 99.9% of the fuel rods in the core do not experience 
transition boiling. Therefore, these changes to the [MCPR] safety 
limit do not affect the probability or consequences of an accident 
[previously evaluated].
    GE's GESTAR-II approved methodology will continue to be 
implemented and has no effect on the probability or consequences of 
any accidents previously evaluated. One exception to GESTAR is that 
the mis-oriented and mis-located bundle events will continue to be 
analyzed as accidents subject to the acceptance criteria in the 
current licensing basis [for GGNS]. The design of the GE11 fuel 
bundles[, to be added to the core to replace Siemens fuel bundles,] 
is such that the bundles are not likely to be mis-oriented or mis-
located and the normal administrative controls will be in effect for 
assuring proper orientation and location. Therefore, the probability 
of a fuel loading error is not increased. This analysis ensures that 
postulated dose releases will not exceed a small fraction (10 
percent) of 10CFR100 [10 CFR Part 100] limits. Therefore, the 
probability or consequences of accidents previously evaluated are 
unchanged.
    II. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The GE 11 fuel to be [added to the core and] used in Cycle 10 
[operation] is of a design compatible with fuel present in the core 
and used in the [current 9th] cycle. [The current core is a mixture 
of GE11 and Siemens fuel bundles. The addition of GE11 to the core 
for the 9th cycle is addressed in Amendment 131 to the license dated 
November 21, 1996.] Therefore, the GE11 fuel will not create the 
possibility of a new or different kind of accident. The proposed 
changes do not involve any new modes of operation, any changes to 
setpoints, or any plant modifications.
    They introduce revised MCPR safety limits that have been proven 
to be acceptable for Cycle 10 operation. Compliance with the 
applicable criterion for incipient boiling transition continues to 
be ensured. The proposed MCPR safety limits do not result in the 
creation of any new precursors to an accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    III. The proposed change does not involve a significant 
reduction in a margin of safety.
    The MCPR safety limits have been evaluated in accordance with 
GE's current cycle-specific methodology to ensure that during normal 
operation and during AOOs, at least 99.9% of the fuel rods in the 
core are not expected to experience transition boiling. Unless 
otherwise approved, GGNS will implement only the NRC-approved 
revisions to GE's GESTAR methodology. This GE methodology is similar 
to those SPC [(Siemens Power Corporation)] reports current listed in 
TS 5.6.5 and it will be applied in a similar, conservative fashion. 
[TS 5.6.5, Core Operating Limits Report, lists the analytical 
methods which are approved by NRC and are used to determine the core 
operating limits for the GGNS core, including the MCPR.] One 
exception to GESTAR is that the mis-oriented and mis-located bundle 
events will continue to be analyzed as accidents subject to the 
acceptance criteria in the current [GGNS] licensing basis. This 
analysis ensures that postulated dose releases will not exceed a 
small fraction (10 percent) of 10CFR100 limits. [The proposed 
changes are to maintain the margin of safety for transition boiling 
in the core.] On this basis, the implementation of this GE 
methodology does not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 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: James W. Clifford, Acting

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

    Date of amendment request: September 25, 1997
    Description of amendment request: The proposed change modifies 
Limiting Condition for Operation (LCO) 3.6.1.2 (Containment Leakage), 
the associated Action, and Surveillance Requirement (SR) 4.6.1.2 in 
Technical Specification (TS) for Waterford Steam Electric Station, Unit 
3 (Waterford 3). The air lock door seal leakage rate acceptance 
criteria in TS 6.15 is being changed from 0.01La to 0.005La. TS 6.15 is 
also being modified to make the terms used in the Containment Leakage 
Rate Testing Program consistent with terms used in the TS. This change 
corrects an error that inadvertently decreased the allowed outage time 
from 24 hours to 1 hour when the containment purge valve or containment 
air lock leakage rates are not within limits. This error was made in 
the Waterford 3 TS change request that was approved in Amendment 124 
for Waterford 3 on April 10, 1997.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No
    The proposed change adds the specific type of containment 
leakage to the Limiting Condition for Operation (LCO), Action, and 
Surveillance Requirement (SR) in the Containment Leakage Technical 
Specification (TS) which results in increasing

[[Page 54873]]

the allowed outage time from 1 hour to 24 hours when the containment 
purge valve or containment air lock leakage rates are not within 
limits. The proposed change revises the air lock door seal leakage 
rate acceptance criteria. Also, the proposed change revises the 
Actions in the Containment Leakage TS to be consistent with the 
Applicability, and revises terms in the Containment Section and 
Administrative Controls Section of the TS to be consistent with the 
Containment Leakage Rate Testing Program. This change will not 
affect the probability of an accident. The containment purge valve 
and air lock leakage rates are not an initiator of any analyzed 
event. This change corrects two errors that were made in the 
Waterford 3 10CFR50 Appendix J, Option B, TS change request that was 
approved in TS Amendment 124. The first error inadvertently 
decreased the allowed outage time from 24 hours to 1 hour when 
either the containment purge valve or containment air lock leakage 
rate acceptance criteria is not met. The second error inadvertently 
increased the acceptance criteria for the air lock door seal 
leakage. The revised air lock door seal leakage rate acceptance 
criteria was never used at Waterford 3. This change also 
administratively changes the Containment Leakage TS Action and terms 
in the TS for consistency.
    The proposed change will not affect the consequences of an 
accident. The amount of leakage from the containment purge valve and 
from the containment air lock will still be included in the overall 
combined containment leak rate. Neither the overall containment 
leakage rate limit nor the Action required to be taken if the 
overall containment leakage rate were exceeded is being changed. The 
Containment Leakage TS Action will be consistent with the 
Applicability and TS 3.0.4 will prohibit entry into Mode 4 (RCS 
[Reactor Coolant System] temperature <ls-thn-eq> 200 deg.F), unless 
the overall containment leakage rate is within limit. The revised 
air lock acceptance criteria was never used. Waterford 3 will 
continue using the more restrictive acceptance criteria which is 
controlled administratively. This proposed change does not affect 
the mitigation capabilities of any component or system, nor does it 
affect the assumptions relative to the mitigation of accidents or 
transients.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No
    The proposed change adds the specific type of containment 
leakage to the LCO, Action, and SR in the Containment Leakage TS. 
This results in increasing the allowed outage time from 1 hour to 24 
hours when the containment purge valve or containment air lock 
leakage rates are not within limits. The proposed change revises the 
air lock door seal leakage rate acceptance criteria. Also, the 
proposed change revises the Actions in the Containment Leakage TS to 
be consistent with the Applicability, and revises terms in the 
Containment Section and Administrative Controls Section of the TS to 
be consistent with the Containment Leakage Rate Testing Program. 
Neither the design nor configuration of the plant, or how the plant 
is operated is being changed due to the addition of the specific 
types of leakage from the Containment Leakage Rate Testing Program, 
corrections made to the air lock door seal leakage rate acceptance 
criteria, or the changes made to make the TS consistent. There has 
been no physical change to plant systems, structures, or components 
nor will these changes reduce the ability of any of the safety-
related equipment required to mitigate anticipated operational 
occurrences or accidents. Therefore, the proposed change will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No
    The proposed change adds the specific type of containment 
leakage to the LCO, Action, and SR in the Containment Leakage TS. 
This results in increasing the allowed outage time from 1 hour to 24 
hours when the containment purge valve or containment air lock 
leakage rates are not within limits. The proposed change revises the 
air lock door seal leakage rate acceptance criteria. Also, the 
proposed change revises the Actions in the Containment Leakage TS to 
be consistent with the Applicability, and revises terms in the 
Containment Section and Administrative Controls Section of the TS to 
be consistent with the Containment Leakage Rate Testing Program. The 
proposed revision to the Action and making the containment leakage 
rate terms consistent are administrative changes that have no 
technical impact on the TS.
    The pre-amendment 124 Waterford 3 TS and NUREG-1432 allowed 
entry into specific Actions with allowed outage times greater than 1 
hour (24 hours) when the air lock and purge valve leakage rate 
acceptance criteria could not be met. This change restores this 
allowed outage time which was inadvertently changed due to an error 
in the TS change request. The increased allowed outage time may 
prevent an unnecessary plant shutdown which is a plant transient. 
Plant shutdowns produce thermal stress on components in the Reactor 
Coolant System and the potential for a plant upset that could 
challenge safety systems. This change decreases the possibility of a 
plant shutdown by replacing the 1 hour allowed outage time with a 24 
hour allowed outage time when the containment purge valve or 
containment air lock leakage is not within limits. Also, the overall 
containment leakage rate limits are not being changed and are 
required to be maintained.
    The revision to the air lock door seal acceptance criteria is a 
more restrictive change to correct an error made by Waterford 3 in 
the TS change request approved in Amendment 124. The less 
restrictive acceptance criteria was never used; Waterford 3 
continued testing to the more restrictive acceptance criteria.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: James W. Clifford, Acting

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

    Date of amendment request: October 1, 1997
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) for the Crystal River Nuclear 
Electric Generating Plant Unit 3 (CR-3). The proposed TS change would 
add a new TS section, 5.6.2.10.4.c. The new section will provide growth 
monitoring criteria for the first span section of tubes in the ``B'' 
Once-Through Steam Generator (OTSG) with pit-like intergranular attack 
(IGA) indications.
    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 purpose of OTSG tube inspection is to identify tubes that 
have a higher potential for in service failure due to degradation 
that results in a reduced ability to withstand normal and upset 
operating conditions. The formal incorporation of specific 
indication growth monitoring and repair criteria is consistent with 
this purpose. Therefore, the probability of an accident previously 
evaluated has not been increased.
    Chapter 14 of the CR-3 Final Safety Analysis Report (FSAR) 
provides an analysis to assess the consequences of a steam generator 
tube rupture event, including the complete severance of a steam 
generator tube. This analyses concluded that CR-3 was sufficiently 
designed to ensure that in the event of a steam generator tube 
rupture, the radiological doses would not exceed the allowable 
limits prescribed by 10 CFR 100. Neither would this result in 
additional tube failures and further degradation of the

[[Page 54874]]

integrity of the reactor coolant pressure boundary. The proposed 
changes do not alter this analysis in any fashion. Therefore, the 
consequences of an accident have not been increased.
    Criterion 2
    Does not Create the Possibility of a New or Different Kind of 
Accident from any Accident Previously Evaluated.
    This change does not alter the design or operation of the OTSGs. 
The incorporation of the proposed requirements is more conservative 
than the existing ITS requirements. Neither the type of inspection 
of OTSG tubes nor the process for performing inspections will be 
changed by this amendment. Therefore, this change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    Criterion 3
    Does Not Involve a Significant Reduction in the Margin of Safety 
as defined in the Bases for any Technical Specifications.
    The previously performed analyses on the effects of OTSG tube 
failures, as reported in the CR-3 FSAR, have demonstrated that 
onsite and offsite consequences are within allowable limits. The 
proposed change incorporates more conservative growth monitoring and 
operational assessment criteria for the ``B'' OTSG first-span pit-
like IGA indications. This change does not result in a significant 
reduction in the margin of safety as defined in the Bases for any 
Technical Specifications.
    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: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC - A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042
    NRC Project Director: Frederick J. Hebdon

North Atlantic Energy Service Corporation, Docket No. 50-443, 
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: September 26, 1997
    Description of amendment request: The proposed amendment would 
separate the requirements for Control Room Air Conditioning from 
Control Room Makeup Air and Filtration as presently contained in 
Technical Specification 3.7.6, ``Control Room Emergency Makeup Air and 
Filtration,'' and its associated BASES. Technical Specification 3.7.6 
now requires that each subsystem of Control Room Emergency Makeup Air 
and Filtration include an OPERABLE emergency filtration unit and air 
conditioning unit. The proposed amendment would separate the 
requirements based on system function. The proposed amendment also 
would increase the allowed outage time for the air conditioning portion 
of the Control Room Air Conditioning Subsystem.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.
    A. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated (10 
CFR 50.92(c)(1)). The proposed changes have no impact on the 
probability of an accident because the control room ventilation 
systems are support systems which have a role in the detection and 
mitigation of accidents but do not contribute to the initiation of 
any accident previously evaluated. Reorganizing the Technical 
Specifications by function is merely an administrative change and 
the change has no impact on the course of any accidents previously 
evaluated since there is no change in the functions provided by the 
subsystems.
    Increasing the allowed outage time to 30 days from 7 days for 
the cooling of recirculated air while one train is inoperable does 
not affect the availability of the second train of air conditioning 
or the actions required if both trains of air conditioning become 
unavailable. Thus, the consequences accidents previously evaluated 
are not increased.
    B. The changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
(10 CFR 50.92(c)(2)) because they do not affect the function of any 
facility structure, system or component, nor do they affect the 
manner by which the facility is operated. The proposed changes do 
not introduce any new failure modes.
    C. The changes do not involve a significant reduction in a 
margin of safety (10 CFR 50.92(c)(3)) because the proposed changes 
do not affect the function of any facility structure, system or 
component, nor do they affect the manner by which the facility is 
operated. Increasing the allowed outage time for the cooling of 
recirculated air while one train is inoperable represents an 
increase in the probability that the air conditioning functions 
could be unavailable. However, the increase does not affect the 
availability of the second train of air conditioning or the actions 
required should both trains of air conditioning become unavailable.
    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: Exeter Public Library, 
Founders Park, Exeter, NH 03833
    Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast 
Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270
    NRC Project Director: Ronald B. Eaton, Acting

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 amendment requests: September 26, 1997
    Description of amendment requests: The proposed amendments would 
revise Technical Specification (TS) 3.4.B, ``Auxiliary Feedwater 
System,'' to provide specific guidance for conducting post-maintenance 
operational testing of the turbine-driven auxiliary feedwater (TDAFW) 
pump and associated system valves to meet operability and limiting 
conditions for operation during unit startup. An additional change is 
proposed to revise Table TS.3.5.2B to permit during Mode 2 the 
bypassing of the auto start feature of the auxiliary feedwater (AFW) 
pumps that results from the trip of both main feedwater pumps when the 
feedwater pumps are not required to be operated.
    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[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Since none of the proposed changes involve a physical change to 
the plant, the mechanisms that could cause a Loss of Normal 
Feedwater have not changed. The probability that a Loss of Normal 
Feedwater will occur is not altered.
    This change still requires that the motor driven AFW Pump and 
associated system valves are operable during Startup Operations. 
Analysis of the Loss of Normal Feedwater transient shows that a 
single AFW Pump provides sufficient AFW flow to prevent any adverse 
conditions in the core. The condition of an inoperable TDAFW Pump is 
already permitted during power operations where the consequences of 
the event would be more severe than during startup. Since there are 
no consequences from the Loss of Normal Feedwater event at power, 
the consequences during startup would still be none, but the margins 
would be larger because; (1) the amount of residual heat generated 
is less because reactor power

[[Page 54875]]

at the start of the event is less and (2) the power history is lower 
resulting in less decay heat.
    Thus, these changes do not involve an increase in the 
probability or consequences of an accident previously analyzed.
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    The proposed changes do not create the possibility of a new or 
different kind of accident previously evaluated because the proposed 
changes do not introduce a new mode of operation or testing, or make 
physical changes to the plant.
    The proposed changes do not alter the design, function, 
operation, or testing of any plant component, therefore the 
possibility of a new or different kind of accident from those 
previously analyzed would not be created by these changes to 
Technical Specifications.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    Margins previously established for the Loss of Normal Feedwater 
event, were analyzed for different initial conditions. The Loss of 
Normal Feedwater event was analyzed for Power Operations. This 
analysis determined that no adverse conditions would occur in the 
core. Since there are no consequences from the Loss of Normal 
Feedwater event at power, the consequences during startup would 
still be none but the margins would be greater because; (1) the 
amount of residual heat generated is less because reactor power at 
the start of the event is less and (2) the power history is lower 
causing less decay heat.
    Therefore, the proposed change does not result in a significant 
reduction in the margin of safety currently established.
    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: 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: John N. Hannon

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

    Date of amendment request: September 3, 1997
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) to revise the number of hours 
operating personnel can work in a normal shift. The proposed amendment 
also contains some administrative changes to the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    A. Establishing operating personnel work hours at, ``an 8 to 12 
hour day, nominal 40 hour week,'' allows normal plant operations to 
be managed more effectively and does not adversely effect 
performance of operating personnel. Overtime remains controlled by 
site administrative procedures in accordance with NRC Policy 
Statement on working hours (Generic Letter 82-12). If 8 hour shifts 
are maintained in part or whole, then acceptable levels of 
performance from operating personnel is assured through effective 
control of shift turnovers and plant activities. No physical plant 
modifications are involved and none of the precursors of previously 
evaluated accidents are affected. Therefore, this change will not 
involve a significant increase in the probability or consequence of 
an accident previously evaluated.
    B. Editorial changes clarify section 6.2.2.g without changing 
the intent or meaning. The proposed change meets the intent of the 
NRC Policy Statement on working hours (Generic Letter 82-12).
    C. Changes to sections 3.10.6.1.a and 3.10.9 do not change the 
intent or meaning of the technical specification sections. 
Clarification to the table notation in section 4.1 related to the 
definition of shift checks to monitor plant conditions will continue 
as intended but are allowed to increase up to at least once per 12 
hours. This increase is consistent with standard industry practice 
as represented by the Standard Technical Specifications (STS), 
Reference 1.
    2. Does the proposed license amendment create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    A. Establishing operating personnel work hours at, ``an 8 to 12 
hour day, nominal 40 hour week,'' allows normal plant operations to 
be managed more effectively and does not adversely effect 
performance of operating personnel. If 8 hour shifts are maintained 
in part or whole, then acceptable levels of performance from 
operating personnel is assured through effective control of shift 
turnovers and plant activities. Overtime remains controlled by site 
administrative procedures in accordance with the NRC Policy 
Statement on working hours (Generic Letter 82-12). No physical 
modification of the plant is involved. As such, the change does not 
introduce any new failure modes or conditions that may create a new 
or different accident. Therefore, operation in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    B. Editorial changes clarify section 6.2.2.g without changing 
the intent or meaning. The proposed change meets the intent of the 
NRC Policy Statement on working hours (Generic Letter 82-12).
    C. Changes to sections 3.10.6.1.a and 3.10.9 do not change the 
intent or meaning of the technical specification sections. 
Clarification to the table notation in section 4.1 related to the 
definition of shift checks to monitor plant conditions will continue 
as intended but are allowed to increase up to at least once per 12 
hours. This increase is consistent with standard industry practice 
as represented by the Standard Technical Specifications (STS), 
Reference 1.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    A. Establishing operating personnel work hours at, ``an 8 to 12 
hour day, nominal 40 hour week,'' allows normal plant operations to 
be managed more effectively and does not adversely effect 
performance of operating personnel. If 8 hour shifts are maintained 
in part or whole, then acceptable levels of performance from 
operating personnel is assured through effective control of shift 
turnovers and plant activities. Overtime remains controlled by site 
administrative procedures in accordance with the NRC Policy 
Statement on working hours (Generic Letter 82-12) and is consistent 
with the Standard Technical Specifications. The proposed change 
involves no physical modification of the plant, or alterations to 
any accident or transient analysis. There is no Basis to section 6 
of the Technical Specifications, and the changes are administrative 
in nature. Therefore, the change does not involve any significant 
reduction in a margin of safety.
    B. Editorial changes clarify section 6.2.2.g without changing 
the intent or meaning. The proposed change meets the intent of the 
NRC Policy Statement on working hours (Generic Letter 82-12).
    C. Changes to sections 3.10.6.1.a and 3.10.9 do not change the 
intent or meaning of the technical specification sections. 
Clarification to the table notation in section 4.1 related to the 
definition of shift checks to monitor plant conditions will continue 
as intended but are allowed to increase up to at least once per 12 
hours. This increase is consistent with standard industry practice 
as represented by the Standard Technical Specifications (STS), 
Reference 1.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. David Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director

[[Page 54876]]

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

    Date of amendment request: September 8, 1997
    Description of amendment request: The proposed amendment would 
revise the f delta I function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously analyzed?

    Response:
    No. The revision to the negative [f delta I] penalty does not 
significantly increase the probability or consequences of an 
accident previously evaluated in the FSAR [Final Safety Analysis 
Report]. This revision does not directly initiate an accident. The 
consequences of accidents previously evaluated in the FSAR are 
unaffected by this proposed change because no change to any 
equipment response or accident mitigation scenario has resulted. 
There are no additional challenges to fission product barrier 
integrity.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response:
    No. The revision to the negative [f delta I] penalty does not 
create the possibility of a new or different kind of accident than 
any accident already evaluated in the FSAR. No new accident 
scenarios, failure mechanisms, or limiting single failures are 
introduced as a result of this proposed change. The proposed 
Technical Specification revision does not challenge the performance 
or integrity of any safety related systems. Therefore, the 
possibility of a new or different kind of accident is not created.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    No. The proposed change to the Technical Specification does not 
involve a significant reduction in a margin of safety. The margin of 
safety associated with the acceptance criteria for any accident is 
unchanged.
    The revision to the negative [f delta I] penalty will have no 
affect on the availability, operability or performance of the safety 
related systems and components and does not affect the plant 
Technical Specification requirements. The revision to the negative 
[f delta I] penalty does require a change to the Technical 
Specifications but does not prevent inspections or surveillances 
required by the Technical Specifications.
    In addition, the revision to the [f delta I] parameters is based 
upon the revised boron dilution rate used to analyze the boron 
dilution transient. Indian Point 3 procedures require the placement 
of one PW [primary water makeup] pump control switch in the pull-out 
position, thus ensuring that only one PW pump is operating.
    The Bases of the Technical Specifications are founded in part on 
the ability of the regulatory criteria being satisfied assuming the 
limiting conditions for operation for various systems. Conformance 
to the regulatory criteria for operation with the revision to the 
negative [f delta I] penalty is demonstrated and the regulatory 
limits are not exceeded. Therefore, the margin of safety as defined 
in the Technical Specifications is not reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601
    Attorney for licensee: Mr. David Blabey, 10 Columbus Circle, New 
York, New York 10019
    NRC Project Director: S. Singh Bajwa, Director

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of amendment request: September 29, 1997
    Description of amendment request: The proposed amendment would 
revise the Ginna Station Improved Technical Specifications (ITS) to 
change the Allowable Value for high steam flow input into limiting 
condition for operation (LCO) Table 3.3.2-1, Function 4.d.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. An increase in 
the high steam flow Allowable Value for LCO Table 3.3.2-1, Function 
4.d does not increase the probability of any analyzed accident nor 
does it increase the likelihood of an inadvertent main steam 
isolation. This function is not explicitly credited in the accident 
analyses. Also, there are three coincident parameters which must be 
reached in order for this function to cause a main steam line 
isolation. It has been demonstrated that the change to the high 
steam flow parameter does not delay the time at which this isolation 
signal would be reached for any analyzed accident since the steam 
flow value is reached much earlier in the accident scenario than the 
other parameters. Therefore, these changes do not involve a 
significant increase in the probability or consequences of an 
accident previously analyzed.
    2. Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes do not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or changes in 
the methods governing normal plant operation. Thus, this change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes do not directly affect any analyzed 
accident analysis. The new isolation times will not be affected for 
analyzed accidents. As such, no question of safety is involved, and 
the change does not involve a significant reduction in a margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005
    NRC Project Director: S. Singh Bajwa, Director

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: August 26, 1997
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 3/4.6.1.3, ``Containment Systems - 
Containment Air Locks,'' TS Bases 3/4.6.1.3, ``Containment Systems - 
Containment Air Locks,'' and TS Bases 3/4.9.4, ``Refueling Operations - 
Containment Penetrations.'' The containment air lock Limiting Condition 
for Operation and Surveillance Requirements would be modified, and the 
associated bases would be changed.
    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

[[Page 54877]]

consideration, which is presented below:
    The Davis-Besse Nuclear Power Station has reviewed the proposed 
changes and determined that a significant hazards consideration does 
not exist because operation of the Davis-Besse Nuclear Power 
Station, Unit No. 1, in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because accident initiators, 
conditions, or assumptions are not affected by the proposed changes, 
which clarify the Technical Specification (TS) Limiting Condition 
for Operation (LCO) for the containment air locks, extend the test 
frequency for the containment air lock interlock mechanisms, and 
modify guidelines relative to the routing of hoses and cables 
through the containment air lock during core alterations or during 
movement of irradiated fuel within the containment.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because the proposed changes do not 
change the source term, containment isolation, or allowable 
releases. The proposed changes do not affect the allowable 
containment leakage rates presently specified in the Technical 
Specifications.
    The proposed change to Surveillance Requirement (SR) 4.6.1.3.c 
to increase the surveillance interval for the air lock interlock 
mechanism to ``at least once per REFUELING INTERVAL'' is justified 
due to the purely mechanical nature of the interlock mechanism, and 
given that the interlock mechanism is not normally challenged when 
the air lock door is used for entry and exit since administrative 
controls require strict adherence to single door opening. Operating 
experience shows that the interlock mechanisms are very reliable. 
Further, the proposed change will allow performance of the 
surveillance under the conditions that apply during a plant outage, 
which is preferable to performance, in part, with the plant at 
power, as is currently necessitated by the present six month 
interval surveillance requirement. Although an interlock mechanism 
failure would not affect air lock sealing capabilities and would 
therefore not directly affect containment integrity, performance of 
the surveillance with the plant at power, when containment integrity 
is required, carries with it the potential for loss of containment 
integrity, should the interlock fail during testing and allow both 
doors to be opened simultaneously. The proposed TS change may result 
in an increased probability that due to the increased [decreased] 
test frequency, an inoperable interlock mechanism could go 
undetected for a longer length of time. However, in the unlikely 
event that as a containment entry is being made, abnormal radiation 
levels inside containment occur, any increase in consequences due to 
a radioactive release as a result of an inadvertent opening of both 
air lock doors (as could be allowed by a failed interlock mechanism 
and assuming violation of administrative controls) is counter-
balanced by the decreased likelihood of similar events occurring 
when the interlock mechanism is tested at power under the current, 
more frequent, test requirement.
    The proposed change to TS Bases 3/4.9.4 to add flexibility in 
routing cable and hoses through the containment personnel air lock 
will not affect the requirement to maintain at least one containment 
personnel air lock door capable of being closed. The analysis 
results for a fuel handling accident inside containment, as 
presented in Section 15.4.7.3 of the DBNPS Updated Safety Analysis 
Report (USAR), are well within the 10 CFR 100 guideline values. 
Since the analysis does not take credit for containment isolation, 
the status of the personnel air lock has no impact on the 
acceptability of the results. Under the proposed change, in the 
event of a fuel handling accident, release of radioactive material 
will continue to be minimized since at least one personnel air lock 
door will remain capable of being closed.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new 
accident initiators or assumptions are introduced by the proposed 
changes. The proposed changes do not involve a change to the plant 
design or operation and, therefore, will not introduce any new or 
different failure modes or initiators.
    3. Not involve a significant reduction in a margin of safety.
    The proposed TS change to SR 4.6.1.3.c to increase the 
surveillance interval for the air lock interlock mechanism will have 
no adverse effect on plant safety based on its good historical 
surveillance and maintenance data, and the reduction in testing at 
power which will occur.
    The analysis results for a fuel handling accident inside 
containment, as presented in the D
    Basis for proposed no significant hazards guideline values. Since 
the analysis does not take credit for containment isolation, the status 
of the personnel air lock has no impact on the acceptability of the 
results. Therefore, the proposed change to TS Bases 3/4.9.4 to add 
flexibility in routing cable and hoses through the containment 
personnel air lock will not reduce the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
Vermont Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: October 11, 1996
    Description of amendment request: The proposed ammendment would 
revise the Vermont Yankee Technical Specifications (TSs) regarding the 
amount of foam concentrate required to support operability of the 
Recirculation Motor Generator (M. G.) Set Foam System as stated in TS 
3.13.G.1 and 3.13.G.2. In both instances, the required amount of foam 
concentrate would be increased from 100 to 150 gallons.
    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 changes proposed herein affect only the amount of foam 
concentrate inventory required to support the operability of the 
Recirculation M. G. Set Foam System and therefore does not modify or 
add any initiating parameters that would significantly increase the 
probability or consequences of any previously analyzed accident.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any previously evaluated:
    These changes involve the upgrade of an existing system using 
standard fire protection components to provide the level of 
protection originally required. An evaluation has been completed to 
ensure that the enhanced spray pattern and increased volume of spray 
does not impact any equipment not previously evaluated and does not 
create any threat of flooding to equipment. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety:
    These changes do not affect any equipment involved in potential 
initiating events or safety limits. Therefore, it is concluded that 
the proposed change does not involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensees 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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301

[[Page 54878]]

    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, N.W., Washington, DC 20037-1128
    NRC Project Director: Ronald B. Eaton, Acting Director

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

    Date of amendment request: September 2, 1997
    Description of amendment request: This license amendment request 
proposes to revise Technical Specification 3.7.1.2, Auxiliary Feedwater 
System, and associated Bases, to add requirements for the essential 
service water (ESW) flowpaths to the turbine-driven auxiliary feedwater 
pump (TDAFWP) and other changes consistent with the technical 
specification conversion application previously submitted. The proposed 
revisions would (a) provide an action and allowed outage time (AOT) for 
inoperability of one of the redundant ESW flowpaths to the TDAFWP, and 
(b) incorporate an action and AOT for inoperability of one of the 
redundant steam flowpaths to the TDAFWP turbine and other changes to 
make the auxiliary feedwater system limiting condition for operation 
(LCO) and actions consistent with those previously submitted.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    ESW Flow Path Required Actions
    This change would provide a 7-day AOT for the ESW supply flow 
paths to the TDAFWP. This would replace administrative controls that 
imposed a 72-hour AOT on ESW flow paths to the TDAFWP.
    The proposed change does not result in any hardware changes or 
changes to operating methodologies. This revision does not affect an 
accident initiator of any analyzed accident since the TDAFWP ESW 
supply only provides flow to equipment required to mitigate the 
consequences of an accident. The revision recognizes that the TDAFWP 
would remain available in most cases for accident mitigation because 
of the low probability of an accident and subsequent equipment 
failure requiring the use of the inoperable ESW supply for the 
TDAFWP. Changing the AOT from 3 days to 7 days would have a 
negligible effect on this small probability. Loss of the AFW 
function would also require the failure of the MDAFWPs [motor-driven 
auxiliary feedwater pumps]. In addition, the CST [condensate storage 
tank] would be OPERABLE in accordance with LCO 3.7.1.3 and would be 
available for use by the TDAFWP for all events except those external 
hazards that represent a hazard to the integrity of the tank itself.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Steam Supply Flow Path Required Actions
    This change would provide a 7-day AOT for the steam supply flow 
paths to the TDAFWP. This would replace an administrative control 
that required the TDAFWP to be declared inoperable without applying 
an AOT. The proposed change does not result in any hardware changes 
or changes to operating methodologies. This revision does not affect 
an accident initiator of any analyzed accident since the TDAFWP 
steam supply only provides power to equipment required to mitigate 
the consequences of an accident. The revision recognizes the low 
probability of an accident requiring the use of the inoperable steam 
supply for the TDAFWP coincident with the failure of the MDAFWPs.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    3. Use of ``Trains'' Instead of ``Pumps and Associated Flow 
Paths'' and Removal of Unnecessary Details
    This change is partially administrative and partially a movement 
of provisions not required to be in the technical specifications to 
other controlled documents. The administrative change does not 
impact initiators of analyzed events or equipment assumed in the 
mitigation of accidents or transient events. The details moved from 
the technical specification would be located in the Bases of the 
technical specification. Since any changes to the Bases will be 
evaluated per the requirements of 10 CFR 50.59, proper controls are 
in place to adequately limit the probability or consequences of an 
accident previously evaluated. Therefore, this change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    4. Twelve Hours to HOT SHUTDOWN
    This change would allow an additional 6 hours to achieve HOT 
SHUTDOWN for the AFW System. The proposed change does not alter the 
plant configuration or operation or function of any safety system. 
Consequently, the change does not increase the probability of an 
accident as defined in accident analysis. The proposed change 
permits a longer time to cooldown to RHR [residual heat removal] 
entry conditions; however, this would not affect the consequences of 
any postulated accidents and is appropriate due to the need to avoid 
any transients while cooling down with a potentially degraded AFW 
System.
    Therefore, the proposed change would have no significant effect 
on the probability or consequences of any previously analyzed 
accidents.
    5. Additional AOT of 10 Days from Discovery of Failure to Meet 
the LCO
    The proposed change imposes more stringent requirements than 
contained in current technical specification. The more stringent 
requirements are imposed to ensure that the OPERABILITY requirements 
for the AFW System are maintained consistent with the safety 
analysis and licensing basis. Therefore, this change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    6. Suspension of LCO 3.0.3
    The proposed change involves clarifying the technical 
specification. The proposed revision involves no technical changes 
to the current technical specification. As such, this change is 
administrative in nature and does not impact initiators of analyzed 
events or assumed mitigation of accidents or transient events. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    1.ESW Flow Path Required Actions
    The proposed change to add a 7-day AOT for the ESW supply flow 
paths does not require physical alteration to any plant system or 
change the method by which any safety-related system performs its 
function.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    2. Steam Supply Flow Path Required Actions
    The proposed change to add a 7-day AOT for the steam supply flow 
paths does not require physical alteration to any plant system or 
change the method by which any safety-related system performs it 
function.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Use of ``Trains'' Instead of ``Pumps and Associated Flow 
Paths'' and Moving of Unnecessary Details
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in controlling parameters. The proposed change will not 
impose any different requirements and adequate control of the 
information moved to the Bases will be maintained. The proposed 
change will not impose any different requirements. Thus, this change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    4. Twelve Hours to HOT SHUTDOWN
    The proposed change does not require physical alteration to any 
plant system or change the method by which any safety-related system 
performs its function. As discussed above, the change does allow 
additional time to complete transfer from the SG [steam generator] 
as the method for heat removal to the RHR System, but does not alter 
the basic methodology.
    Therefore, the proposed change would not create the possibility 
of a new or different kind of accident.

[[Page 54879]]

    5. Additional AOT of 10 Days from Discovery of Failure to Meet 
the LCO
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in controlling parameters. The proposed change does 
impose different (more restrictive) requirements. However, these 
changes remain consistent with assumptions made in the safety 
analysis regarding system OPERABILITY. Thus, this change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    6. Suspension of LCO 3.0.3
    The proposed change clarifies an implied requirement from 
current technical specifications and does not involve a physical 
alteration of the plant (no new or different type of equipment will 
be installed) or changes in controlling parameters. The proposed 
change will not impose any different requirements. Thus, this change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    1. ESW Flow Path Required Actions
    The proposed change to add a 7-day AOT for the ESW flow paths 
does not change any accident analysis assumptions, initial 
conditions or results. Consequently, it does not have an effect on 
margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    2. Steam Supply Flow Path Required Actions
    The proposed change to add a 7-day AOT for the steam supply flow 
paths does not change any accident analysis assumptions, initial 
conditions or results. Consequently, it does not have an effect on 
margin of safety.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    3. Use of ``Trains'' Instead of ``Pumps and Associated Flow 
Paths'' and Removal of Unnecessary Details
    The proposed change will not reduce a margin of safety because 
it has no impact on the design basis or safety analysis. In 
addition, the requirements to be transposed from the technical 
specification to the Bases are the same as the current technical 
specification. Since any future changes to these requirements in the 
Bases will be evaluated per the requirements of 10 CFR 50.59, proper 
controls are in place to maintain an appropriate margin of safety. 
Therefore, the changes do not involve a significant reduction in a 
margin of safety.
    4. Twelve Hours to HOT SHUTDOWN
    The proposed change does not alter the basic regulatory 
requirements or change any accident analysis assumptions, initial 
conditions or results.
    Therefore, the proposed change would have no significant adverse 
effect on margins of safety.
    5. Additional AOT of 10 Days from Discovery of Failure to Meet 
the LCO
    The imposition of more stringent requirements on AOT would 
increase the margin of plant safety by providing additional 
requirements to maintain AFW System OPERABILITY.
    The change is consistent with the safety analysis and licensing 
basis. Therefore, this change does not involve a reduction in a 
margin of safety.
    6. Suspension of LCO 3.0.3
    The proposed change will not reduce a margin of safety because 
it has no impact on the design basis or safety analysis. This change 
is administrative in nature. As such, no question of safety is 
involved.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

Yankee Atomic Electric Company, Docket No. 50-029, Yankee Nuclear 
Power Station, Franklin County, Massachusetts

    Date of amendment request: September 5, 1997 (Accession No. 
9709100106)
    Description of amendment request: The proposed technical 
specification (TS) changes are needed to permit removal of spent 
nuclear fuel from the Spent Fuel Pit storage racks into a combined 
storage/shipping cask and to enable handling of the cask components and 
other hardware by the Yard Area Crane. Specific TS changes are needed 
for minimum water coverage over spent fuel, shielding for personnel 
exposure, increased loads carried over the fuel, addition of 
restrictions for load paths over spent fuel and changes to the 
appropriate TS bases.
    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 changes provide for an alternate method of providing 
protection of the spent fuel and spent fuel pit (SFP) from heavy 
loads that must be transported over the SFP. The method chosen, that 
is, providing a single-failure-proof overhead crane, is considered 
an acceptable method as stated in Regulatory Guide 1.13, ``Spent 
Fuel Storage Facility Design Basis,'' and NUREG-0612, ``Control of 
Heavy Loads at Nuclear Power Plants.'' The Defueled Technical 
Specification 3.1.2 requirement for five (5) feet of water above the 
top of the fuel assemblies for fuel traveling in the SFP is provided 
for personnel protection (ALARA). This protection is provided by the 
shielding afforded by the shipping and/or transfer cask system. The 
cask handling crane will comply with the single-failure-proof crane 
design requirements of NUREG-0554, ``Single Failure-Proof Cranes for 
Nuclear Power Plants,'' and meet the criteria specified in NUREG-
0612. In addition, design controls and administrative controls will 
be maintained to prevent handling of the shipping and/or transfer 
cask over spent fuel in the SFP. As such, these changes will not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated. NUREG-0612, Section 
5, provides direction for providing an adequate level of defense-in-
depth for handling of heavy loads near spent fuel and safe shutdown 
systems. The single-failure-proof overhead crane design is presented 
as an acceptable method of providing the proper margin of safety for 
handling of heavy loads. By upgrading the cask handling crane to a 
single-failure-proof design and meeting the requirements presented 
in Sections 5.1.1 and 5.1.6 of NUREG-0612 (for safe load path, 
procedures, crane operator training and qualification, special 
lifting devices, lifting devices that are not specially designed, 
and crane inspection, testing, and maintenance) a sufficient level 
of defense-in-depth is provided to ensure that a load drop is not a 
credible event. As such, there is no increase in the probability or 
consequence of an accident previously evaluated as a result of the 
heavy load changes. A fuel handling incident is a currently analyzed 
event; dropping of a fuel assembly over the spent fuel within the 
transfer cask is similar to dropping of a fuel assembly over spent 
fuel in the SFP. The design basis fuel handling event analysis 
bounds these events, so there is no 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 defense-in-depth philosophy 
provided by the single-failure-proof crane load handling sysem design, 
and compliance with the requirements specified in Sections 5.1.1 and 
5.1.6 of NUREG-0612 provide assurance that for a credible single 
failure of the crane load handling system, the system will still be 
able to perform its safety function. This provides assurance that a 
load drop accident is not a credible event. As such, no new or 
different kind of accident will be created from any accident previously 
evaluated.
     3. Involve a significant reduction in a margin of safety. The 
proposed changes implement the guidelines of NUREG-0612 and 
Regulatory Guide 1.13. YAEC is implementing an acceptable alternate 
method of ensuring the safe handling of heavy loads

[[Page 54880]]

over the SFP. This method provides a defense-in-depth approach for 
handling of heavy loads over the SFP and maintains the margin of 
safety consistent with that of the current requirements. Further 
protection is provided by the prohibition of these additional heavy 
loads from travel over the spent fuel assemblies in the SFP racks. 
The use of a single-failure-proof crane and associaed lifting 
devices provide an increased margin of safety that ensure that a 
load drop event is not credible and is considered an adequate 
alternate for the additional area added to the safe load path. The 
use of a limit switch to prevent movement of the prohibited cask 
handling crane loads from movement beyond the safe load path, 
provides an additional margin of safety, that was previously 
provided by the steel framing at the southern edge of the SFP 
superstructure roof opening. The single-failure-proof crane and 
defense-in-depth design ensure that a load drop is not a credible 
event, assuring that the margin of safety is not reduced.
    Based on the above considerations, it is concluded that there is 
reasonable assurance that the operation of Yankee Nuclear Power 
Station consisent with the proposed changes will not endanger the 
health and safety of the public.
    The proposed change has been reviewed by the Plant Operations 
Review Committee and the Nuclear Safety Audit and Review Committee.
    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: Greenfield Community College, 1 College 
Drive, Greenfield, Massachusetts 01301
    Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One 
International Place, Boston, Massachusetts 02110-2624
    NRC Project Director: Seymour H. Weiss

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.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: September 12, 1997
    Brief description of amendment: The proposed amendment involves a 
revision to the Emergency Diesel Generator protective relaying scheme 
at CR3, as described in the Final Safety Analysis Report Chapter 8.
    Date of publication of individual notice in the Federal Register: 
September 30, 1997 (62 FR 51165).
    Expiration date of individual notice: October 30, 1997
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal River, Florida 34428

Notice Of Issuance Of Amendments To Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

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

    Date of application for amendment: March 24, 1995, as supplemented 
by letters dated September 10, 1995, and March 22, 1996.
    Brief description of amendment: The amendment would change the 
technical specifications (TS) to (1) reflect the applicable portions of 
NUREG-1432, ``Standard Technical Specifications Combustion Engineering 
Plants,'' (2) implement the recommendations of Generic Letter (GL) 93-
05, ``Line Item Technical Specification Improvements to Reduce 
Surveillance Requirements for Testing During Plant Operation,'' and (3) 
implement the recommendations of GL 94-01, ``Removal of Accelerated 
Testing and Specific Reporting Requirements for Emergency Diesel 
Generators.'' The purpose of the proposed amendment is to increase 
emergency diesel generator (EDG) reliability by reducing stresses on 
EDG caused by unnecessary testing. The associated Bases are also 
updated.
    Date of issuance: October 6, 1997
    Effective date: October 6, 1997, to be implemented within 120 days 
of date of issuance.
    Amendment Nos.: Unit 1 - 114; Unit 2 - 107; Unit 3 - 86
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29870) The September 10, 1995, and March 22, 1996, supplemental letters 
provided additional clarifying information and did not change the 
original no significant hazards consideration. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
October 6, 1997. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004

[[Page 54881]]

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: March 28, 1996, as supplemented 
November 20, 1996, and July 31, 1997.
    Brief description of amendments: The amendments reduce the 
moderator temperature coefficient limit shown on Technical 
Specification Figure 3.1.1-1. This proposed change is necessary to 
support changes in the safety analyses made to accommodate a larger 
number of plugged steam generator tubes for future operating cycles.
    Date of issuance: October 2, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 222 and 198
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Registe for amendment: February 
21, 1997
    Brief description of amendment: This amendment adds a specific time 
limit to Technical Specification Table 3.3-3 to place an inoperable 
refueling water storage tank level channel in a bypassed condition.
    Date of issuance: September 30, 1997
    Effective date: September 30, 1997
    Amendment No.: 74
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17225) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 30, 1997. 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. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: July 1, 1997
    Brief description of amendments: The amendments revise Technical 
Specification Table 3.3.7.1-1, ``Radiation Monitoring 
Instrumentation,'' to require two channels to be operable per trip 
system as opposed to two per intake. This change reflects a 
modification to the design of the instrumentation logic to satisfy 
single failure requirements. The amendments also revise the associated 
action statement to clarify system logic wording.
    Date of issuance: October 9, 1997
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 121 and 106
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45455). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 9, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348

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 application for amendments: May 1, 1997
    Brief description of amendments: The amendments clarify the load 
value for the emergency diesel generator to be equal to or greater than 
the largest single load and revise the frequency and voltage 
requirements during the performance of the test.
    Date of issuance: October 7, 1997
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 178 and 176
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33121). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 7, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van 
Buren County, Michigan

    Date of application for amendment: January 10, 1996, as 
supplemented February 20, 1997
    Brief description of amendment: The amendment revises the Technical 
Specifications for the containment emergency escape air lock test 
requirements. Concurrently, the Commission has also granted an 
exemption to certain requirements of 10 CFR Part 50, Appendix J, 
relating to the testing of the emergency escape air lock, to the extent 
that leakage rate testing is not necessary after opening the emergency 
escape air lock doors for post-test restoration or seal adjustment.
    Date of issuance: September 30, 1997
    Effective date: September 30, 1997
    Amendment No.: 177
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 26, 1997 (62 
FR 8795) The February 20, 1997, letter provided clarifying information 
within the scope of the original application and did not change the NRC 
staff's initial proposed no significant hazards considerations 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 30, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423 Consumers Energy Company, Docket No. 
50-255, Palisades Plant, Van Buren County, Michigan
    Date of application for amendment: December 6, 1995, as 
supplemented October 18 1996, January 10 and June 27, 1997
    Brief description of amendment: The amendment deletes crane 
operation and movement of heavy loads requirements and their bases from 
the technical specifications. The requirements have been incorporated 
into the Palisades Operating Requirements Manual (ORM). The ORM has 
been incorporated by reference into the Palisades Final Safety Analysis 
Report, assuring that future changes to the crane and heavy loads 
requirements will be subject to the provisions of 10 CFR 50.59.
    Date of issuance: October 2, 1997
    Effective date: October 2, 1997
    Amendment No.: 178
    Facility Operating License No. DPR-20: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 17, 1996 (61 FR 
37298) The October 18, 1996, January 10 and June 27, 1997, letters 
provided clarifying information within the scope of the original 
application and did not change the staff's initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated October 2, 1997. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423

[[Page 54882]]

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

    Date of amendment request: September 5, 1997 (NRC-97-0107)
    Description of amendment request: The amendment revises the 
Technical Specifications by adding a special test exception to allow 
reactor coolant temperatures up to 212 degrees Fahrenheit during 
hydrostatic or inservice leak testing while in Operational Condition 4 
without entering Operational Condition 3. The amendment also makes 
related changes to the Index, Table 1.2, ``Operational Conditions,'' 
and the Bases to incorporate the reference to the proposed special test 
exception. Date of issuance: September 30, 1997
    Effective date: September 30, 1997, with full implementation within 
45 days
    Amendment No.: 114
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications and Bases.
    Date of initial notice in Federal Register: September 30, 1997 (62 
FR The Commission's related evaluation of the amendment, finding of 
exigent circumstances, consultation with the State of Michigan, and 
final determination of no significant hazards considerations are 
contained in a Safety Evaluation dated September 30, 1997 No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161

Duke Energy Corporation, et al., Docket No. 50-413, Catawba Nuclear 
Station, Unit 1, York County, South Carolina

    Date of application for amendment: May 8, 1997, as supplemented by 
letter dated September 10, 1997
    Brief description of amendment: The amendment revises Section 3/
4.1.2 of the Technical Specifications to permit a one-time natural 
circulation test during Mode 3.
    Date of issuance: October 9, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment No.:  162
    Facility Operating License No. NPF-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30631) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 9, 1997. No significant 
hazards consideration comments received: NoLocal Public Document Room 
location: York County Library, 138 East Black Street, Rock Hill, South 
Carolina 29730

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and 
Entergy Operations, Inc., Docket No. 50-458, River Bend Station, 
Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: August 5, 1997, as supplemented August 
15, 1997
    Brief description of amendment: The amendment revises the Technical 
Specifications to increase the two recirculation loop Minimum Critical 
Power Ratio (MCPR) safety limit to 1.13 and the single recirculation 
loop MCPR safety limit to 1.14.
    Date of issuance: October 8, 1997
    Effective date: October 8, 1997
    Amendment No.: 99
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45456) The August 15, 1997, submittal provided clarifying information 
that did not change the initial no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 8, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803

Florida Power and Light Company, et al., Docket No. 50-389, St. 
Lucie Plant, Unit No. 2, St. Lucie County, Florida

    Date of application for amendment: August 1, 1997
    Brief description of amendment: Revises the Technical 
Specifications (TS) to extend the surveillance interval for the 
Engineered Safety Features Actuation System to a refueling interval on 
a staggered test basis.
    Date of Issuance: October 2, 1997
    Effective Date: October 2, 1997
    Amendment No.: 90
    Facility Operating License No. NPF-16: Amendment revised the TS.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45457) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 2, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Indian River Community College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596

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

    Date of application for amendment: August 14, 1997, as supplemented 
September 9, 19, and 24, 1997
    Brief description of amendment: The amendment revises the TMI-1 
Technical Specifications which decreases the maximum allowable dose 
equivalent iodine-131 limit in the reactor primary coolant from 1.0 
uCi/gm to 0.35 uCi/gm.
    Date of Issuance: October 2, 1997
    Effective Date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 204
    Facility Operating License No. NPF-50: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45459) The supplemental letters did not affect the initial no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated October 2, 1997. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105

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: April 10, 1997
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) by relocating the TS surveillance requirement for 
attaining a negative pressure in the enclosure building, addressing 
operability, deleting the definition for enclosure building integrity, 
modifying enclosure building access opening requirements, and making 
editorial changes for clarification and consistency. The TS Bases are 
also updated to reflect the proposed changes including the need to 
maintain the integrity of the enclosure building and to support 
previously approved laboratory testing requirements for charcoal filter 
sample testing.
    Date of issuance: September 30, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 208

[[Page 54883]]

    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 7, 1997 (62 FR 
24987) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 30, 1997. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385

Northeast Nuclear Energy Company, et al., Docket No. 50-423, 
Millstone Nuclear Power Station, Unit No. 3, New London County, 
Connecticut

    Date of application for amendment: July 18, 1997
    Brief description of amendment: The amendment adds a new Technical 
Specification and associated Bases to address the operability of the 
steam generator atmospheric relief bypass valves.
    Date of issuance: October 2, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 151
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 13, 1997 (62 FR 
43370) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 2, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

PECO Energy Company, Public Service Electric and Gas Company, 
Delmarva Power and Light Company, and Atlantic City Electric 
Company, Docket No. 50-278, Peach Bottom Atomic Power Station, Unit 
No. 3, York County, Pennsylvania

    Date of application for amendment: January 17, 1995, as 
supplemented by letters dated March 30, 1995, July 2, 1996, February 
28, 1997, and September 22, 1997
    Brief description of amendment: The amendment revised the technical 
specifications to support the replacement of the Source Range and 
Intermediate Range Monitors with the Wide Range Neutron Monitoring 
System.
    Date of issuance: September 30, 1997
    Effective date: As of its date of issuance and is to be implemented 
upon completion of Unit 3 Modification P00271.
    Amendment No.: 224
    Facility Operating License No. DPR-56: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 6, 1995 (62 FR 
29885) The March 30, 1995, July 2, 1996, February 28, 1997, and 
September 22, 1997, supplemental letters 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 September 30, 1997. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105

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: April 14, 1997
    Brief description of amendment: The amendment revises Appendix A, 
Section 6 of the James A. FitzPatrick Technical Specifications. These 
changes will enable the Safety Review Committee to review rather than 
audit plant staff performance by deleting the plant staff performance 
audit requirements from Section 6.5.2.9.b and incorporating a plant 
staff performance review requirement in Section 6.5.2.8. Additionally, 
this amendment application replaces the position title of Vice 
President Regulatory Affairs and Special Projects with Director 
Regulatory Affairs and Special Projects.
    Date of issuance: October 3, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 240
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 13, 1997 (62 FR 
43374) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 3, 1997. 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

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: June 19, 1997, as supplemented 
by letters dated July 30 and 31, 1997
    Brief description of amendment: This amendment changes TS 
4.1.3.1.2, ``Control Rod Operability;'' TS 3.1.3.6, ``Control Rod Drive 
Coupling;'' TS 3.1.3.7, ``Control Rod Position Indication;'' TS 
3.1.4.1, ``Rod Worth Minimizer;'' TS 3/4.1.4.2, ``Rod Sequence Control 
System;'' TS 3/4.10.2, ``Special Test Exceptions - Rod Sequence Control 
System;'' the Bases for TS 2.2.1.2, ``Average Power Range Monitor;'' 
the Bases for TS 3/4.1.4, ``Control Rod Program Controls;'' and the 
Bases for TS 3/4.10.2, ``Rod Sequence Control System.'' The changes 
eliminate the Rod Sequence Control System (RSCS) Limiting Condition for 
Operation and Surveillance Requirements from the TSs and reduce the Rod 
Worth Minimizer low power setpoint to 10% from 20%. Changes to other 
sections of the TSs delete reference to the RSCS from the TSs and 
incorporate additional requirements necessary to support the 
elimination of the RSCS.
    Date of issuance: September 30, 1997
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No.: 105
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications and the License.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45462) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 30, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston 
County, Alabama

    Date of amendments request: May 28, 1997
    Brief description of amendments: The amendments revise the 
Technical Specifications to clarify that testing of

[[Page 54884]]

each shared emergency diesel generator (EDG), 1-2A and 1C, to comply 
with surveillance requirement 4.8.1.1.2.e is only required once per 5 
years on a per EDG basis, not on a per unit basis.
    Date of issuance: October 1, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 129, 122
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33135) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 1, 1997. No significant 
hazards consideration comments received: No.Local Public Document Room 
location: Houston-Love Memorial Library, 212 W. Burdeshaw Street, Post 
Office Box 1369, Dothan, Alabama 36302

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of 
Georgia, City of Dalton, Georgia, Docket No. 50-321, Edwin I. Hatch 
Nuclear Plant, Unit 1, Appling County, Georgia

    Date of application for amendment: May 9, 1997, as supplemented 
September 19, 1997
    Brief description of amendment: The amendment revises the minimum 
critical power ratio safety limits for a mixed core of GE9B/GE12/GE13 
fuel for Cycle 18 operation.
    Date of issuance: October 8, 1997
    Effective date: Prior to the restart from the Hatch Unit 1 outage 
currently scheduled to begin October 1997.
    Amendment No.: 209
    Facility Operating License No. DPR-57: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40857) The September 19, 1997, submittal provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
October 8, 1997. No significant hazards consideration comments 
received: No.Local Public Document Room location: Appling County Public 
Library, 301 City Hall Drive, Baxley, Georgia 31513

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of 
Georgia, City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, 
Edwin I. Hatch Nuclear Plant, Units 1 and 2, Appling County, 
Georgia

    Date of application for amendments: May 9, 1997, as supplemented 
September 3, 1997
    Brief description of amendments: The amendments revise the 
applicability requirements for the Rod Block Monitor (RBM) to require 
that the RBM be operable whenever reactor thermal power is greater than 
or equal to 29 percent of rated thermal power.
    Date of issuance: October 8, 1997
    Effective date: As of the date of issuance to be implemented prior 
to Unit 1 startup from the fall 1997 refueling outage for Unit 1; and 
implemented within 30 days from issuance for Unit 2.
    Amendment Nos.: 210, 151
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40857) The September 3, 1997, submittal 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 October 8, 1997. 
No significant hazards consideration comments received: No.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

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: August 21, 1996, as 
supplemented by letters dated March 17, March 27, April 3, and July 15, 
1997 (TS 96-07)
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) by revising the as-found setpoint 
tolerance band for the pressurizer Code safety relief valves and the 
main steam Code safety relief valves from plus or minus one percent to 
plus or minus three percent.
    Date of issuance: September 29, 1997
    Effective date: September 29, 1997
    Amendment Nos.: 229 (Unit 1), 220 (Unit 2)
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise TS.
    Date of initial notice in Federal Register: October 9, 1996 (61 FR 
52969) The March 17, March 27, April 3, and July 15, 1997, letters 
provided clarifying information that did not change the initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 29, 1997. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library 1001 Broad Street, Chattanooga, Tennessee 37402

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: August 14, 1997 (TSCR 199)
    Brief description of amendments: The amendments revise TS 15.4.2.B. 
``In-Service Inspection and Testing of Safety Class Components Other 
than Steam Generator Tubes,'' to modify item 2 by deleting the 
reference to TS 15.4.4 and referencing the Containment Leakage Rate 
Testing Program; TS 15.6.12.A.1, ``Containment Leakage Rate Testing 
Program,'' to eliminate the one-time requirement for Unit 2 Type A 
testing since the testing has been completed; and TS Bases 15.4.4 to 
delete the specific bases for containment purge valve testing and to 
delete a reference that is no longer used. Date of issuance: September 
29, 1997Effective date: September 29, 1997, with full implementation 
within 45 days
    Amendment Nos.:  181 and 185
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45466) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 29, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams 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: July 29, 1997
    Brief description of amendment: The amendment changes the wording 
of Action Statement 5a to Technical Specification Table 3.3-1, 
``Reactor Trip System Instrumentation.'' This action statement 
prescribes a set of actions to

[[Page 54885]]

 be accomplished when a source range neutron detector is inoperable 
with the plant shutdown. The proposed wording change will clarify the 
times and order in which these actions are to be performed.
    Date of issuance: September 29, 1997
    Effective date: September 29, 1997, to be implemented within 30 
days from the date of issuance.
    Amendment No.: 111
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45467) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 29, 1997. 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

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

    Date of amendment request: September 6, 1997
    Brief description of amendment: This amendment allows the testing 
of certain contacts in the emergency diesel generator load sequencer to 
be done with the unit at power (Mode 1) and provides an additional 24 
hours to the time allowed by TS 4.0.3 to complete the testing.
    Date of issuance: October 7, 1997
    Effective date: October 7, 1997
    Amendment No.: 112
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes (62 FR 49261 dated September 19, 
1997). The notice provided an opportunity to submit comments on the 
Commission's proposed no significant hazards consideration 
determination. No comments have been received. The notice also provided 
for an opportunity to request a hearing by October 20, 1997, but 
indicated that if the Commission makes a final no significant hazards 
consideration determination any such hearing would take place after 
issuance of the amendment. The Commission's related evaluation of the 
amendment, finding of exigent circumstances, and final determination of 
no significant hazards consideration are contained in a Safety 
Evaluation dated October 7, 1997.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    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 15th day of October 1997.
    For the Nuclear Regulatory Commission
Elinor G. Adensam,
Acting DirectorDivision of Reactor Projects - III/IV, Office of Nuclear 
Reactor Regulation
[Doc. 97-27877 Filed 10-21-97; 8:45 am]
BILLING CODE 7590-01-F