[Federal Register Volume 63, Number 116 (Wednesday, June 17, 1998)]
[Notices]
[Pages 33103-33119]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-16012]


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NUCLEAR REGULATORY COMMISSION


Biweekly Notice; Applications and Amendments to Facility 
Operating Licenses Involving no Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from May 22,

[[Page 33104]]

1998, through June 5, 1998. The last biweekly notice was published on 
June 3, 1998 (63 FR 30261).

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 Administration 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 July 17, 1998, 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 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,

[[Page 33105]]

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. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 
1 and 2, Rock Island County, Illinois

    Date of application for amendment request: May 18, 1998.
    Description of amendment request: Change various technical 
specification (TS) values to conservatively reflect design values. 
These TS values affect: (1) 125/250 volts direct current (Vdc) 
electrolyte temperature; (2) control rod drive accumulator pressure; 
(3) standby liquid control solution temperature; (4) ultimate heat sink 
minimum water level; (5) shutdown suppression chamber level (Quad 
Cities only); and (6) degraded voltage setpoint (Quad Cities only).
    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:
    Does the change involve a significant increase in the probability 
or consequences of an accident previously evaluated?
    The change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
proposed changes to certain Technical Specification acceptance values 
are conservative and serve to ensure operability of equipment important 
to safety. By ensuring equipment availability, the probability or 
consequences of an accident previously evaluated are not increased. In 
addition, the proposed changes have no impact on any initial condition 
assumptions for accident scenarios. Onsite or offsite dose consequences 
resulting from an event previously evaluated are not affected by this 
proposed amendment request.
    Accordingly, there is no significant change in the probability or 
consequences of an accident previously evaluated.
    Does the change create the possibility of a new or different kind 
of accident from any accident previously evaluated?
    The proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. The 
proposed license amendment provides changes in certain Technical 
Specification values to restore margin and ensure equipment 
operability. Each proposed change is conservative with respect to 
current requirements. The proposed amendment does not involve any plant 
physical changes that would create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Therefore, the proposed amendment does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed change does not involve a significant reduction in a 
margin of safety. In fact, the proposed changes restore margin and 
ensure equipment operability. Since the changes maintain the necessary 
level of system reliability, they do not involve a significant 
reduction in the margin of safety.
    Therefore, 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: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Stuart A. Richards.

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

    Date of application for amendment: May 2, 1998, as supplemented May 
21, and 23 (three letters), 1998.
    Brief description of amendment: This amendment changed Technical 
Specification (TS) 3/4.6.2, ``Protective Instrumentation,'' and its 
associated Bases to reflect modifications to the initiation 
instrumentation for the Control Room Air Treatment System. It also 
changed TS 3.2.4a, ``Reactor Coolant Activity,'' and added an 
additional condition to the operating license.
    Date of issuance: May 23, 1998.
    Effective date: As of the date of issuance to be implemented prior 
to resumption of power operation.
    Amendment No.: 161.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes (63 FR 27601 dated May 19, 1998. 
The notice recognized the existence of exigent circumstances pursuant 
to 10 CFR 50.91(a)(6) and provided an opportunity to submit comments on 
the Commission's proposed no significant hazards consideration 
determination. The notice published May 19, 1998, also provided for an 
opportunity to request a hearing by June 1, 1998 (this will be 
corrected to June 18, 1998, by a notice to be published in the near 
future), 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. Subsequent to publishing 
the notice, and due to schedule improvements which have occurred at the 
plant, the Commission has determined that the amendment should be 
issued on an emergency basis pursuant to 10 CFR 50.91(a)(5). The 
Commission's related evaluation of the amendment, finding of emergency 
circumstances, consultation with the State of New York, and final no 
significant hazards consideration determination are contained in a 
Safety Evaluation date May 23, 1998.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State

[[Page 33106]]

University of New York, Oswego, New York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
    NRC Project Director: S. Singh Bajwa.

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

    Date of amendment request: May 14, 1998.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) for the Reactor Protection 
System (RPS) and the Engineered Safety Features Actuation System 
(ESFAS) instrumentation by restricting the time most RPS and ESFAS 
actuation channels can be in the bypass position to 48 hours. The 
current TSs have no time limit. The proposed amendment would also 
modify the TS action requirements and the channel calibration 
requirements for the loss of turbine load reactor trip function, and 
the channel calibration requirements for the wide range logarithmic 
neutron flux monitors; add a note to exclude the neutron detectors from 
the channel calibration requirements; correct a reference to a TS 
surveillance requirement; and correct errors that have been identified.
    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 evaluated.
    The proposed change to restrict the time most of the reactor 
protection or engineered safety feature actuation channels can be in 
the bypass position to 48 hours, from an indefinite period of time, has 
no effect on the design of the Reactor Protection System (RPS) or the 
Engineered Safety Feature Actuation System (ESFAS), and does not affect 
how these systems operate. In addition, this will minimize the 
susceptibility of these systems to the remote possibility of fault 
propagation between channels. The pressurizer high pressure reactor 
protection channels will not be required to be placed in the tripped 
condition after 48 hours. A failed pressurizer high pressure channel 
will be allowed to remain in the bypassed condition for up to 30 days. 
If the failed pressurizer high pressure channel was placed in the 
tripped condition, and then a high failure of another pressurizer high 
pressure channel occurred, the reactor would trip and both pressurizer 
power operated relief valves (PORVs) would open, resulting in an 
undesired loss of primary coolant. Limiting the time that a failed 
pressurizer high pressure reactor protection channel can be in bypass 
to 30 days will minimize the risk of the inadvertent opening of both 
PORVs, as well as the risk associated with fault propagation between 
channels. These systems will still function as designed to mitigate 
design basis accidents. Therefore, this change does not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    The proposed change to increase the time a second RPS or ESFAS 
channel can be removed from service (from 2 hours to 48 hours), 
provided one of the inoperable channels is placed in the tripped 
condition, has no effect on the design of the RPS or ESFAS and does not 
affect how these systems operate. These systems will still function as 
designed to mitigate design basis accidents.
    However, one of the proposed changes will allow two pressurizer 
pressure reactor protection channels to be removed from service (one 
channel in the tripped condition and one channel in the bypassed 
condition) for 48 hours instead of the current 2 hour time limit. With 
a pressurizer pressure channel in the tripped condition, the high 
failure of a second pressurizer pressure channel would initiate a 
reactor trip, open both pressurizer PORVs, and cause an undesired loss 
of primary coolant. Thus, this change will increase the probability of 
occurrence of a previously evaluated accident (FSAR [Final Safety 
Analysis Report] Section 14.6.1--Inadvertent Opening of a Pressurized 
Water Reactor Pressurizer Pressure Relief Valve). However, since this 
configuration will only be allowed for an additional 46 hours, the 
increase in the probability of occurrence of a previously evaluated 
accident will be limited to an acceptable value. Therefore, this change 
does not significantly increase the probability or consequences of an 
accident previously evaluated.
    The proposed change to apply a more restrictive action statement to 
the loss of turbine load reactor trip function has no effect on the 
design of this trip function and does not affect how this trip function 
operates. Also, this trip function is not assumed to operate to 
mitigate any design basis accident.
    Therefore, this change does not significantly increase the 
probability or consequences of accident previously evaluated.
    The proposed change to require a channel calibration every 18 
months for the loss of turbine load reactor trip function and for the 
wide range logarithmic neutron flux monitors has no effect on the 
design of either the loss of turbine load reactor trip function or the 
wide range logarithmic neutron flux monitors. Also, neither of these 
are assumed to operate to mitigate any design basis accident. 
Therefore, this change does not significantly increase the probability 
or consequences of an accident previously evaluated.
    The proposed change to exclude the neutron detectors from the 
channel calibration requirement has no effect on the design of the 
neutron detectors and has no significant effect on how these detectors 
operate. The detectors are passive devices with minimal drift. In 
addition, slow changes in the sensitivity of the linear power range 
flux detectors is compensated for by performing the daily calorimetric 
calibration and the monthly calibration using the incore detectors. 
These detectors will still function as designed to mitigate design 
basis accidents. Therefore, this change does not significantly increase 
the probability or consequences of an accident previously evaluated.
    The proposed change to correct the surveillance requirement 
referenced in an action statement has no effect on the design of the 
ESFAS and does not affect how this system operates. The ESFAS will 
still function as designed to mitigate design basis accidents. 
Therefore, this change does not significantly increase the probability 
or consequences of an accident previously evaluated.
    The proposed change to add a reference to the reactor coolant pump 
low speed reactor trip function to a note that states this trip may be 
bypassed when [less than] 5 [percent] power, and that the bypass must 
be automatically removed when [greater than or equal to] 5 [percent] 
power will not effect this reactor trip function. This bypass 
capability currently exists in the design of the Millstone Unit No. 2 
RPS, and is the same bypass feature referenced for the reactor coolant 
flow low reactor trip function. Both of these reactor trip functions 
provide protection for a reduction in RCS [Reactor Coolant System] 
flow. The addition of this note will not result in any technical change 
to the Millstone Unit No. 2 RPS. The RPS will continue to function as 
before. Therefore, this change does not significantly increase the 
probability or consequences of an accident previously evaluated.

[[Page 33107]]

    The proposed change to correct the power level high trip setpoint 
on Technical Specification Page 2-4 will not result in any change to 
the actual plant setpoint for this RPS trip function. As a result of 
this proposed change, the setpoint listed on Page 2-4 will agree with 
the setpoint previously approved by the NRC, and currently used by the 
RPS. The change has no effect on the design of the RPS and does not 
affect how this system operates. Therefore, this change does not 
significantly increase the probability or consequences of an accident 
previously evaluated.
    The information added to the Bases of the Technical Specifications 
to provide a discussion of how the RPS and ESFAS are affected by the 
proposed changes, the effect the action statements have on the 
operation of the RPS and ESFAS, and to discuss the impact of 
surveillance testing on RPS operability will have no effect on 
equipment operation. The RPS and ESFAS will continue to function as 
designed to mitigate design basis accidents. Therefore, this change 
does not significantly increase the probability or consequences of an 
accident previously evaluated.
    Thus, this License Amendment Request does not impact the 
probability of an accident previously evaluated nor does it involve a 
significant increase in the consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not alter the plant configuration (no new 
or different type of equipment will be installed) or require any new or 
unusual operator actions. They do not alter the way any structure, 
system, or component functions and do not alter the manner in which the 
plant is operated. The proposed changes do not introduce any new 
failure modes. They will not alter assumptions made in the safety 
analysis and licensing basis. The RPS and the ESFAS will still function 
as designed to mitigate design basis accidents.
    Therefore, these changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will not reduce the margin of safety since 
they have no impact on any safety analysis assumption. The proposed 
changes do not decrease the scope of equipment currently required to be 
operable or subject to surveillance testing, nor do the proposed 
changes affect any instrument setpoints or equipment safety functions.
    The effectiveness of Technical Specifications will be maintained 
since the changes will not alter the operation of any RPS or ESFAS 
function. In addition, most of the changes are consistent with the 
Calvert Cliffs RPS and ESFAS Technical Specifications mode provided in 
Enclosure 3 of the NRC correspondence dated April 16, 1981 (R. A. Clark 
letter to W. G. Counsil, Evaluation of the Reactor Protection System 
Inoperable Channel Condition at Millstone Nuclear Power Station, Unit 
No. 2, dated April 16, 1981) and the new, improved Standard Technical 
Specifications (STS) for Combustion Engineering plants (NUREG-1432).
    Therefore, there is no 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: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Deputy Director: Phillip F. McKee.

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: June 25, 1997.
    Description of amendment request: The proposed amendment would 
change the Indian Point 3 Technical Specifications to allow the use of 
zirconium alloy or stainless steel filler rods in fuel assemblies to 
replace failed or damaged fuel rods.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Consistent with the criteria of 10 CFR 50.92, the enclosed 
application is judged to involve no significant hazards based on the 
following information:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident previously 
analyzed?
    Response: The proposed changes modify the technical specification 
only to the extent that the reconstitution is recognized as acceptable 
under limited circumstances. Reconstitution is limited to substitution 
of zirconium alloy or stainless steel filler rods, and must be in 
accordance with approved applications of fuel rod configurations. 
Although these changes permit reconstitution to occur without the need 
for a specific technical specification change, use of an approved 
methodology is required prior to its application. Since the changes 
will allow substitution of filler rods for leaking, potentially leaking 
rods or damaged rods, the changes may actually reduce the radiological 
consequences of an accident. It is noted that the specific changes 
requested in this letter have previously been found acceptable by the 
NRC in GL [Generic Letter] 90-02, Supplement 1. For these reasons, we 
conclude that the changes will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    (2) Does the proposed license amendment create the possibility of a 
new or different kind of accident from any previously evaluated?
    Response: The proposed changes will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated because they will only affect the assembly configuration and 
can only be implemented if demonstrated to meet current plant 
requirements in accordance with an NRC-approved methodology. The other 
aspects of plant design, operation limitations, and responses to events 
will remain unchanged. It is noted that the changes have previously 
been determined acceptable by the NRC in GL 90-02, Supplement 1.
    (3) Does the proposed amendment involve a significant reduction in 
a margin of safety?
    Response: The proposed change will not involve a reduction in a 
margin of safety because the changes can only be implemented if 
demonstrated to meet current plant requirements in accordance with an 
NRC-approved methodology. It is noted that the changes have previously 
been determined acceptable by the NRC in GL 90-02, Supplement 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.

[[Page 33108]]

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.

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

    Date of amendment request: April 28, 1998.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.4.2.1 to replace the plus or 
minus 1 percent setpoint tolerance limit for safety/relief valves 
(SRVs) with a plus or minus 3 percent setpoint tolerance limit. In 
addition, the proposed amendment would revise TS 4.4.2.2 to state that 
all SRVs must be certified to be within plus or minus 1 percent of the 
TS setpoint prior to returning the valves to service.
    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 changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed TS revisions involve: (1) no significant hardware 
changes; (2) no significant changes to the operation of any systems or 
components in normal or accident operating conditions; and (3) no 
changes to existing structures, systems, or components. Therefore these 
changes will not increase the probability of an accident previously 
evaluated.
    These proposed changes were developed in accordance with the 
provisions contained in an NRC Safety Evaluation Report, dated 3/8/93, 
for the ``BWR Owners Group Inservice Pressure Relief Technical 
Specification [Revision] Licensing Topical Report'', NEDC-31753P as 
described in General Electric report NEDC-32511P, ``Safety Review for 
Hope Creek [Generating Station] Safety/Relief Valve Tolerance 
Analyses''. Since the plant systems associated with these proposed 
changes will still be capable of: (1) meeting all applicable design 
basis requirements; and (2) retain the capability to mitigate the 
consequences of accidents described in the HC [Hope Creek] UFSAR 
[Updated Final Safety Analysis Report], the proposed changes were 
determined to be justified. Therefore, these changes will not involve a 
significant increase in the 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.
    Establishment of the [plus or minus] 3 [percent] SRV setpoint 
tolerance limit will not adversely impact the operation of any safety 
related component or equipment. Since the proposed changes involve: (1) 
no significant hardware changes; (2) no significant changes to the 
operation of any systems or components; and (3) no changes to existing 
structures, systems, or components, there can be no impact on the 
occurrence of any accident. These proposed changes were developed in 
accordance with the provisions contained in an NRC Safety Evaluation 
Report, dated 3/8/93, for the ``BWR Owners Group Inservice Pressure 
Relief Technical Specification [Revision] Licensing Topical Report'', 
NEDC-31753P as described in General Electric report NEDC-32511P, 
``[Safety Review for Hope Creek Generating Station] Safety/Relief Valve 
Tolerance Analyses''. Furthermore, there is no change in plant testing 
proposed in this change request which could initiate an event. 
Therefore, these changes will 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.
    Establishment of the [plus or minus] 3 [percent] SRV setpoint 
tolerance limit will not adversely impact the operation of any safety 
related component or equipment. General Electric analyses performed for 
Hope Creek and contained in General Electric report NEDC-32511P, 
``[Safety Review for Hope Creek Generating Station] Safety/Relief Valve 
Tolerance Analyses,'' concluded that there is no significant impact on 
fuel thermal limits, no significant impact on safety related systems, 
structures or components, and no significant impact on the accident 
analyses associated with the proposed changes. Therefore, the changes 
contained in this request do not result in 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: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: May 8, 1998.
    Description of amendment request: The proposed amendments would 
change the Vogtle Electric Generating Plant (VEGP) Technical 
Specification (TS) 5.5.7, ``Reactor Coolant Pump Flywheel Inspection 
Program,'' to provide an exception to the examination requirements of 
Regulatory Position C.4.b of Regulatory Guide (RG) 1.14, Revision 1, 
August 1975.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The safety function of the RCP [reactor coolant pump] flywheel is 
to provide sufficient rotational inertia to ensure reactor coolant flow 
through the core during coastdown following a loss of offsite power and 
subsequent reactor trip. FSAR [Final Safety Analysis Report] Chapter 15 
analysis for a complete loss of forced reactor coolant flow 
demonstrates that the reactor trip together with the flow sustained by 
the inertia of the RCP impeller will be sufficient to prevent the most 
limiting fuel assembly from exceeding the DNBR [departure from nucleate 
boiling ratio] limits.
    The maximum mechanical loading on the RCP motor flywheel results 
from overspeed following a LOCA [loss-of-coolant accident]. The 
analysis presented in WCAP-14535A demonstrates that the revised 
inspection program proposed by this license amendment will ensure the 
integrity of the RCP flywheels will be maintained.

[[Page 33109]]

    Based upon the findings of WCAP-14535A, the ability of the RCP 
flywheel to perform its intended safety function will be unaffected by 
the license amendment and the FSAR Chapter 15 analysis will remain 
valid. Therefore, these proposed changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed license amendment will not change the physical plant 
configuration nor the modes of operation of any plant equipment. Based 
upon the results of WCAP-14535A, no new failure mechanism will be 
introduced by the revised RCP flywheel inspection program. Therefore, 
the proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The operating limits and functional capabilities of the affected 
systems, structures, and components will be unchanged by the proposed 
amendment. The results of the RCP flywheel inspections performed 
throughout the industry and at VEGP have identified no indications 
which would affect its integrity. As presented in WCAP-14535A, detailed 
stress analysis and risk assessments have been completed with the 
results indicating that there would be no change in the probability of 
failure for RCP flywheels if all inspections were eliminated. 
Therefore, these changes do not involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia.
    NRC Project Director: Herbert N. Berkow.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: May 6, 1998.
    Description of amendment request: The proposed amendment would 
replace the two percent penalty addressed in surveillance requirement 
(SR) 3.2.1.2(a) with a burnup-dependent factor to be specified in the 
Watts Bar Core Operating Limits Report (COLR). Specifically, the 
following changes are being proposed:
    1. SR 3.2.1.2(a) and its associated BASES will have the phrase ``by 
a factor of 1.02'' deleted and replaced with the phrase ``by the 
appropriate factor specified in the COLR.''
    2. Technical Specification (TS) Section 5.9.5(b)(3) would be 
updated to reference the revised WCAP (10216-P-A, Revision 1A, 1994) 
that details the analytical methods utilized for the new penalty 
factor.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    A. The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The proposed change involves only the manner in which the penalty 
factors for FQ(Z) would be specified (i.e., burnup-dependent 
factor specified in the Core Operating Limits Report [COLR] versus a 
constant factor specified in the TS). This is simply used to account 
for the fact that FQ C(Z) may increase between 
surveillance intervals. These penalty factors are not assumed in any of 
the initiating events for the accident analyses. Therefore the proposed 
change will have no effect on the probability of any accidents 
previously evaluated. The penalty factors specified in the COLR will be 
calculated using NRC-approved methodology and will continue to provide 
an equivalent level of protection as the existing TS requirement. 
Therefore, the proposed change will not affect the consequences of any 
accident previously evaluated.
    B. The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed change does not involve a physical alteration to the 
plant (no new or different kind of equipment will be installed) or 
alter the manner in which the plant would be operated. Thus, this 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    C. The proposed amendment does not involve a significant reduction 
in a margin of safety.
    The proposed change will continue to ensure that potential 
increases in FQ C(Z) over a surveillance interval 
will be properly accounted for. The penalty factors will be calculated 
using an NRC-approved methodology. Therefore, the proposed change will 
not involve a reduction in margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: May 1, 1998.
    Description of amendment request: The proposed amendment would make 
several editorial changes to the Administrative Controls section of the 
Technical Specifications. The changes include revisions due to 
organizational changes, quality assurance changes, editorial changes, 
and typographical corrections.
    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 the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The administrative change proposed herein will have no effect on 
plant hardware, plant design, safety limit setting or plant system 
operation and therefore do[es] not modify or add any initiating 
parameters that would significantly increase the probability or 
consequences of any previously analyzed accident. The proposed 
amendment changes the reference to the VYNPS QA program and makes other

[[Page 33110]]

administrative changes, such as title changes and correction/
clarification of errors. Therefore, there is no increase in the 
probability or consequence of an accident previously evaluated.
    2. Will the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    This change does not affect any equipment nor does it involve any 
potential initiating events that would create any new or different kind 
of accident. The proposed change involves [ ] wording changes in the 
Technical Specifications identifying the name of the QA program and 
makes other administrative changes, such as title changes and 
corrective/clarification of errors. Therefore no new or different kind 
of accident has been introduced.
    3. Will the proposed changes involve a significant reduction in a 
margin of safety?
    This change does not affect any equipment involved in potential 
initiating events or safety limits. The proposed change has no 
significant impact on margin of safety, as it is comprised of only 
administrative changes.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Project Director: Cecil O.Thomas.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: September 1, 1995, as supplemented April 
8, 1996, April 22, 1996, April 23, 1996, November 18, 1997, February 9, 
1998, March 25, 1998 and May 5, 1998. This notice supersedes the 
Federal Register notice of September 27, 1995 (60 FR 49949)
    Description of amendment request: The originally (September 1, 
1995) proposed changes to the Technical Specifications (TS) would 
permit a single outage of up to 14 days for each emergency diesel 
generator (EDG) once every 18 months in order to perform preventive 
maintenance. The amended request will permit a single outage of up to 
14 days for each EDG for any reason; TS change to incorporate a 
Configuration Risk Management Program (CRMP) in the Administrative 
Section in the TS, in support of the previous submittal for the 14-day 
Allowed Outage Time (AOT) for the EDGs and would permit an increase in 
the TS maintenance interval of the EDG from 18 to 24 months, based on 
the recommendation from the EDG owners group (Fairbanks Morse Owners 
Group).
    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. Specifically, operation of North Anna Power Station in 
accordance with the proposed Technical Specification changes will not:
    a. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    A probabilistic safety analysis (PSA) has been performed which 
demonstrates that a 14-day AOT for each EDG, results in a small change 
in core damage frequency assuming adequate compensatory measures are in 
place. The compensatory measures include requirements that the other 
EDGs, off-site power supply, and the alternate A.C. diesel (AAC DG) be 
operable whenever the action statement is entered.
    The effect of the proposed change has been calculated to be an 
increase in core damage frequency of approximately 1 E-6 per year from 
the baseline core damage frequency of 4.1 E-5. Considering that credit 
was not taken for the AAC DG previously in the IPE nor was the AAC DG 
specified in Technical Specifications, the proposed changes remain 
bounded by the core damage frequency identified in the Individual Plant 
Examination.
    Credit for the AAC DG was previously not taken nor was the AAC DG 
previously included in the Technical Specifications. Furthermore, the 
probabilistic safety analysis (PSA) demonstrates that the increase in 
core damage frequency due to extending the EDG AOT of a 14-day period 
is not significant as long as the AAC DG is operable to act as a source 
of emergency power to replace the EDG. The period of time during which 
the EDG is unavailable is short enough to limit the impact of using the 
manually operated AAC DG as a replacement for the automatically 
operated EDG.
    The plant design and operation are not changed by the incorporation 
of a CRMP into the Administrative Section of Technical Specifications. 
Further, with the proposed change to the preventive maintenance 
interval, the EDG reliability remains adequate to perform its function 
of supporting accident mitigation equipment with emergency electrical 
power.
    Therefore, neither the probability of occurrence nor the 
consequences of an accident or malfunction of equipment important to 
safety previously evaluated in the safety analysis report are increased 
due [to] the proposed changes to permit a 14-day allowed outage time 
and a 24 month preventive maintenance interval for the EDGs.
    b. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No new initiators are defined as a result of a review of the PSA 
model. The proposed Technical Specifications changes only modify the 
AOT of an EDG. The UFSAR [Updated Final Safety Analysis Report] 
accidents are analyzed assuming that the EDG is the worst single 
failure. This assumption is more severe than the proposed Technical 
Specifications changes, which [replace] the EDG with the AAC DG. 
Similarly, the PSA performed to evaluate the proposed Technical 
Specifications changes considered all of the initiating events defined 
for the PSA performed for the Individual Plant Examination. No new 
initiators were defined as a result of a review of the PSA model.
    Adding the CRMP and changing the EDG preventive maintenance 
interval in the Technical Specifications does not change any method of 
operation or create any new modes of operation or accident precursors.
    Therefore, it is concluded that no new or different kind of 
accident or malfunction from any previously evaluated has been or will 
be created by the proposed changes to permit a 14-day allowed outage 
time and a 24 month preventive maintenance interval for the EDGs.
    c. The proposed Technical Specifications changes do not result in a 
reduction in margin of safety as defined in the basis for any Technical 
Specifications.
    The PSA was performed to evaluate the concept of a one-time outage. 
The results of the analyses show a small change in the core damage 
frequency. As described above the proposed Technical Specifications 
changes only modify the AOT of an EDG. Thus, operation with slightly 
increased EDG unavailability due to maintenance is acceptable given the 
operability of the AAC DG and the other EDG.
    Incorporating the CRMP and changing the EDG preventive maintenance 
interval in the Technical Specifications

[[Page 33111]]

does not affect any accident analysis assumptions or change any 
Technical Specifications criteria.
    Therefore, the margin of safety is not changed.
    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: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Gordon E. Edison, Acting.

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

    Date of amendment request: May 7, 1998.
    Description of amendment request: Technical Specification 5.4, 
``Fuel Storage,'' would be changed to increase the allowable mass of 
uranium-235, per axial centimeter, for fuel storage in new fuel and 
spent fuel storage racks. This change will allow use of new Siemens 
heavy fuel assemblies.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change was reviewed in accordance with the provisions 
of 10 CFR 50.92 to show no significant hazards exist. The proposed 
change will not:
    (1) Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The mass of the fuel assembly is increased by a small amount (30 
pounds, or 2.4%), from that of the fuel assemblies now in the core. 
Even with this increase, the load on the fuel handling equipment is 
still well within design limits. Therefore, the probabilities of a fuel 
handling accident inside containment (FHAIC) and the fuel handling 
accident outside containment (FHAOC) are not changed.
    The total core mass, with Siemens heavy fuel, is less than that 
assumed in the original plant safety analysis. The proposed change does 
not alter the plant configuration, operating set points, or overall 
plant performance. The probability of other accidents is therefore not 
changed.
    Attachment 4 (of the application) shows that the consequences of a 
fuel handling accident or a large break loss of coolant accident are 
not significantly affected.
    Any changes in the nuclear properties of the reactor core that may 
result from a higher mass of fuel U235 per axial centimeter 
will be analyzed and shown to meet acceptance criteria in the 
appropriate reload analysis, which would be completed prior to use.
    (2) Create the possibility of a new or different kind of accident 
from any previously evaluated.
    As discussed above, the only safety issue significantly affected by 
the proposed change is the criticality analysis of the spent fuel 
storage racks and new fuel storage racks. Since it has been 
demonstrated that keff remains below the keff 
acceptance criteria, no new or different accident would be created 
through the use of fuel with up to 56.067 grams of U235 per 
axial centimeter at the Kewaunee Nuclear Power Plant.
    The proposed change does not alter the plant configuration, 
operating set points, or overall plant performance and therefore does 
not create a new or different kind of accident from any accident 
previously evaluated.
    (3) Involve a significant reduction in the margin of safety.
    The criticality analysis in Reference 3 (of the application) 
demonstrates that adequate margins to criticality can be maintained 
with up to 56.067 grams of U235 per axial centimeter stored 
in either the new fuel storage racks or the spent fuel storage racks.
    The bounding cases of the analysis demonstrate that keff 
remains less than 0.95 in the spent fuel storage racks and the new fuel 
storage racks if flooded with unborated water. The bounding cases of 
the analysis also demonstrate that keff remains less than 
0.98 in the new fuel storage racks if moderated by optimally misted 
moderator. Therefore, the 56.067 grams of U235 per axial 
centimeter limit is acceptable for storage in both the new fuel storage 
racks and the spent fuel storage racks.
    Any changes in the nuclear properties of the reactor core that may 
result from a higher mass of fuel U235 per axial centimeter 
will be analyzed in the appropriate reload analysis to ensure 
compliance with applicable reload considerations and requirements.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Acting Project Director: Ronald R. Bellamy.

Wisconsin Electric Power Company, Docket No. 50-301, Point Beach 
Nuclear Plant, Unit 2, Town of Two Creeks, Manitowoc County, Wisconsin

    Date of amendment request: May 15, 1998 (NPL-98-0303).
    Description of amendment request: The proposed amendment revises 
the schedule for implementing the boron concentration changes related 
to the planned conversion of Unit 2 to 18-month fuel cycles.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendment will not result in a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes are administrative only. There are no physical 
changes to the facility or its operation. All Limiting Conditions of 
Operation, Limiting Safety System Settings, and Safety Limits specified 
in the Technical Specification remain unchanged. Additionally, there 
are no changes in the Quality Assurance Program, Emergency Plan, 
Security Plan, and Operator Training and Requalification Program. 
Therefore, an increase in the probability or consequences of an 
accident previously evaluated cannot occur.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes are administrative only. No changes to the 
facility structures, systems and components or their operation will 
result. The design and design basis of the facility remain unchanged. 
The plant safety analyses remain current and accurate. No new or 
different failure mechanisms are introduced. Therefore,

[[Page 33112]]

the possibility of a new or different kind of accident from any 
accident previously evaluated is not introduced.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendment does not involve a significant reduction in a 
margin of safety.
    The proposed [amendment is] administrative only. All safety margins 
established through the design and facility license including the 
Technical Specifications remain unchanged. Therefore, all margins of 
safety are maintained.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cynthia A. Carpenter.

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.

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

    Date of amendment request: May 20, 1998 (NRC-98-0099).
    Description of amendment request: The proposed amendment would 
modify the scram discharge volume (SDV) vent and drain valve action 
requirements to be consistent with those contained in NUREG-1433, 
Revision 1, ``Standard Technical Specifications General Electric 
Plants, BWR/4.''
    Detroit Edison is requesting that this license amendment request be 
processed in an exigent manner in accordance with 10 CFR 50.91(a)(6) 
because delay in granting this amendment could lead to a plant 
shutdown.
    Date of publication of individual notice in Federal Register: May 
28, 1998 (63 FR 29254).
    Expiration date of individual notice: Comments: June 11, 1998; 
hearing: June 29, 1998.
    Local Public Document Room location: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.

Duke Energy Corporation, Docket Nos. 50-413 and 50-414, Catawba Nuclear 
Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: May 22, 1998.
    Description of amendment request: The proposed amendments would 
revise Surveillance Requirement Section 4.4.3.3 of the Technical 
Specifications. Section 4.4.3.3 currently requires that the emergency 
power supply for the pressurizer heaters be demonstrated OPERABLE at 
least once per 18 months by manually transferring power from the normal 
to the emergency power supply. The licensee proposed to delete the 
``manual'' requirement because the power supply transfer at the unit 
was designed to be automatic. The proposed requirement is to verify 
that required pressurizer heaters are capable of being powered from an 
emergency power supply once per 18 months.
    Date of publication of individual notice in Federal Register: June 
1, 1998 (63 FR 29759).
    Expiration date of individual notice: July 1, 1998.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

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

    Date of application for amendment: May 2, 1998.
    Brief description of amendment: The amendment changes the Technical 
Specifications 3/4.6.2, ``Protective Instrumentation,'' to reflect 
modifications to the initiation instrumentation for the Control Room 
Air Treatment system.
    Date of publication of individual notice in Federal Register: May 
19, 1998 (63 FR 27601).
    Expiration date of individual notice: June 18, 1998.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: May 15, 1998 (two letters).
    Brief description of amendment: The amendment changes 
administrative sections of the Technical Specifications to reflect a 
restructuring of upper management organization.
    Date of publication of individual notice in Federal Register: June 
2, 1998 (63 FR 30026).
    Expiration date of individual notice: July 2, 1998.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of amendment request: May 12, 1998.
    Brief description of amendment request: These amendments relocate 
certain requirements related to fire protection from the TSs to the 
Updated Final Safety Analysis Report. The TS sections to be relocated 
are: 3/4.3.7.9, Fire Detection Instrumentation; 3/4.7.6, Fire 
Suppression Systems; 3/4.7.7, Fire Rated Assemblies; and 6.2.2e, Fire 
Brigade Staffing. The amendments also replace License Condition 2.C.(6) 
for Unit 1 and License Condition 2.C.(3) for Unit 2. These amendments 
are consistent with the guidance of NRC Generic Letter (GL) 86-10, 
``Implementation of Fire Protection Requirements,'' and GL 88-12, 
``Removal of Fire Protection Requirements from Technical 
Specifications.''
    Date of publication of individual notice in Federal Register: May 
21, 1998 (63 FR 28010).
    Expiration date of individual notice: June 22, 1998.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

[[Page 33113]]

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

    Date of amendment request: March 31, 1997, as supplemented June 18, 
1997, October 10, 1997, October 20, 1997, November 11, 1997, December 
22, 1997, January 15, 1998, January 27, 1998, March 30, 1998, April 23, 
1998, and April 27, 1998.
    Brief description of amendment request: The proposed amendment 
would revise the Ginna Station Improved Technical Specifications to 
reflect a planned modification to the spent fuel pool storage racks.
    Date of publication of individual notice in Federal Register: May 
12, 1998 (63 FR 26213). This notice supersedes the March 31, 1997, 
application published on April 30, 1997 (62 FR 23502).
    Expiration date of individual notice: June 11, 1998.
    Local Public Document Room Location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

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.

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: January 31, 1997, as 
supplemented February 13, February 28, March 25, April 16, August 19, 
and September 29, 1997, January 22, March 17, April 8, April 21, 1998, 
and May 22, 1998.
    Brief description of amendments: The amendments revise the TS for a 
reduction of the total reactor coolant system flow limit from 370,000 
gallons per minute (gpm) to 340,000 gpm in support of increased steam 
generator tube plugging.
    Date of issuance: May 23, 1998.
    Effective date: As of the date of issuance Unit 1 to be implemented 
within 60 days and Unit 2 prior to startup from the spring 1999 
refueling outage.
    Amendment Nos.: 228 and 202.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 26, 1997 (62 
FR 8780).
    The February 13, February 28, March 25, April 16, August 16, and 
September 29, 1997, January 22, March 17, April 8, and April 21, 1998, 
and May 22, 1998, letters provided clarifying information that did not 
change the initial proposed no significant hazards consideration.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated May 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendments: July 18, 1997.
    Brief description of amendments: The amendments revise the listed 
design suppression chamber temperature of 200 deg.F to 220 deg.F and 
the listed total water and steam volume of the reactor coolant system 
from 18,670 cubic feet to 18,320 cubic feet, respectively.
    Date of issuance: May 27, 1998.
    Effective date: May 27, 1998.
    Amendment Nos.: 195 and 225.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
revise the facility's Technical Specifications.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45454).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 27, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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

    Date of application for amendments: June 12, 1997, as supplemented 
February 2, 1998. The February 2, 1998, submittal contained clarifying 
information only and did not change the initial proposed no significant 
hazards consideration or expand the scope of the original Federal 
Register Notice.
    Brief Description of amendments: The amendments consist of changes 
to the Technical Specifications (TS) to revise the Limiting Condition 
for Operation of the TS to limit the drywell average air temperature 
rather than primary containment air temperature. Additionally, the 
amendments require that the drywell average air temperature be 
maintained less than or equal to 150  deg.F during plant operation. The 
current primary containment average temperature limit is 135  deg.F.
    Date of issuance: May 28, 1998.
    Effective date: May 28, 1998.
    Amendment Nos.: 196 and 226.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45454) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 28, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

[[Page 33114]]

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

    Date of application for amendments: October 28, 1997
    Brief Description of amendments: The amendments revise certain 
instrumentation allowable values in the current technical 
specifications to the Improved Technical Specifications format.
    Date of issuance: May 28, 1998.
    Effective date: May 28, 1998.
    Amendment Nos.: 197 and 227. Facility Operating License Nos. DPR-71 
and DPR-62: Amendments change the Technical Specifications.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68304)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 28, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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

    Date of amendment request: November 15, 1995.
    Brief description of amendment: The amendments modify the channel 
functional test interval in the Technical Specifications Surveillance 
Requirements for the Electrical Protective Assemblies in the Reactor 
Protection System.
    Date of issuance: May 29, 1998.
    Effective date: May 29, 1998.
    Amendment No.: 198 and 228.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34887).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 29, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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

    Date of amendment request: November 16, 1994, as supplemented by 
letters dated February 14, 1995, and April 9, 1998.
    Brief description of amendment: The amendments change the Technical 
Specifications (TS) for Units 1 and 2 to revise the basis for removing 
the suppression chamber water temperature monitoring instrumentation 
requirements from the TS. This change is being processed in parallel 
with the Improved Technical Specification conversion.
    Date of issuance: May 29, 1998.
    Effective date: May 29, 1998.
    Amendment Nos.: 199 and 229.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
497)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 29, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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

    Date of application for amendments: April 4, 1996, as supplemented 
January 24, 1997, March 31, 1997, April 2, 1997, April 14, 1997, March 
24, 1998, and May 20, 1998.
    Brief Description of amendments: The amendments modify Technical 
Specifications (TS) 3.0.4, 4.0.3, and 4.0.4, and their associated Bases 
in accordance with the guidance provided in Generic Letter 87-09, 
``Sections 3.0 and 4.0 of the Standard Technical Specifications (STS) 
on the Applicability of Limiting Conditions for Operation and 
Surveillance Requirements.''
    Date of issuance: June 2, 1998.
    Effective date: June 2, 1998.
    Amendment Nos.: 200 and 230.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: July 17, 1996 (61 FR 
37297).
    The supplemental submittals contained clarifying information only, 
and did not change the initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297

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

    Date of amendment request: April 30, 1997, as supplemented October 
28, 1997, and May 15, 1998.
    Brief description of amendment: The amendments revise surveillance 
requirements 4.7.2.b.2 and 4.7.2.c to require testing of the control 
room emergency ventiliation system charcoal adsorber in accordance with 
the American Society for Testing and Material D3803-1989, ``Standard 
Test Method for Nuclear-Grade Activated Carbon.''
    Date of issuance: June 2, 1998.
    Effective date: June 2, 1998.
    Amendment Nos.: 201 and 231.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40846).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297

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

    Date of application for amendments: April 3, 1998
    Brief description of amendments: The amendments revise the 
specified total volume of the condensate storage tank capacity 
requirements from 150,000 gallons to 228,200 gallons to ensure the Core 
Spray System requirement of 50,000 gallons.
    Date of issuance: June 5, 1998.
    Effective date: June 5, 1998.
    Amendment Nos.: 202 and 232.

[[Page 33115]]

    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
revise the facility's Technical Specifications.
    Date of initial notice in Federal Register: May 6, 1998 (63 FR 
25103).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 5, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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

    Date of application for amendment: October 29, 1997.
    Brief description of amendment: This amendment changes Technical 
Specifications (TS) 3.8.1.1.a.3, 3.8.1.1.b.4, and 3.8.1.1.d.2 by 
eliminating the plant shutdown requirements in these TS, and allowing 
the applicable redundant feature TS to direct the plant shutdown when 
required.
    Date of issuance: May 22, 1998.
    Effective date: May 22, 1998.
    Amendment No.: 78.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68305).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 22, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

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

    Date of application for amendment: April 24, 1998, as supplemented 
by letter dated May 15, 1998.
    Brief description of amendment: This amendment revises TS 3.3.2, 
``Engineered Safety Features Actuation System Instrumentation,'' such 
that surveillance of the undervoltage relays may be performed without 
entry into TS 3.0.3. Specifically, the change modifies Table 3.3-3 to 
allow operation with more than one channel of the emergency bus 
undervoltage relays inoperable.
    Date of issuance: June 3, 1998.
    Effective date: June 3, 1998.
    Amendment No.: 79.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 4, 1998 (63 FR 
24574).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 3, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Commonwealth Edison Company, Docket Nos. STN 50-454 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 application for amendments: September 24, 1997.
    Brief description of amendments: The amendments revise the 
surveillance frequency for the turbine throttle valves and the turbine 
governor valves from monthly to quarterly.
    Date of issuance: May 26, 1998.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 103 and 93.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 11, 1998 (63 FR 
11917).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 26, 1998.
    No significant hazards consideration comments received: No.
    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.

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

    Date of application for amendment: January 28, 1998 (NRC-98-0006), 
as supplemented on March 10, 1998 (NRC-98-0036).
    Brief description of amendment: The amendment revises technical 
specification surveillance requirement 4.4.3.2.2.a for the leak rate 
test of the pressure isolation valves, extending it from the current 
18-month interval to a 24-month interval.
    Date of issuance: May 28, 1998.
    Effective date: May 28, 1998, with full implementation within 90 
days.
    Amendment No.: 118.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 25, 1998 (63 
FR 9598).
    The March 10, 1998, supplement requested a change in the 
implementation period. This information was within the scope of the 
original Federal Register notice and did not change the staff's initial 
proposed no significant hazards considerations determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated May 28, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.

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

    Date of application for amendment: November 22, 1995 (NRC-95-0124), 
as supplemented February 19, April 19, May 3, June 12, and December 4, 
1996, January 30 and August 7, 1997, and April 27 and May 22, 1998.
    Brief description of amendment: The amendment revises technical 
specification (TS) 3.8.1.1 to change the emergency diesel generator 
(EDG) allowed outage time from 3 to 7 days and add a requirement to 
verify that combustion turbine-generator 11-1 is available prior to 
removing an EDG from service. In addition, in accordance with draft 
staff guidance for risk-informed amendments, a section is added to the 
Administrative Controls Section of the TS describing the licensee's 
configuration risk management program. The associated Bases are also 
revised. The November 22, 1995, submittal also requested changes to the 
testing and reporting requirements for the EDGs. These aspects were 
addressed in Amendment No. 107 to the TS issued on June 20, 1996. The 
staff's action on the licensee's request is now complete.
    Date of issuance: June 2, 1998.
    Effective date: June 2, 1998, with full implementation within 60 
days.
    Amendment No.: 119.
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7550) with a supplemental notice on May 1, 1998 (63 FR 24195).

[[Page 33116]]

    The February 19, April 19, May 3, June 12, and December 4, 1996, 
August 7, 1997, and May 22, 1998, submittals provided clarifying 
information within the scope of the Federal Register notices and did 
not change the staff's initial proposed no significant hazards 
considerations determinations.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.

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

    Date of application for amendments: March 17, 1998, as supplemented 
May 14, 1998.
    Brief description of amendments: These amendments revise Action 34 
of technical specification (TS) Table 3.3-3, ``Engineered Safety 
Feature Actuation System Instrumentation.'' Action 34 is applicable to 
Functional Units 6.b., ``Grid Degraded Voltage (4.16 kV Bus),'' and 
6.c., ``Grid Degraded Voltage (480 v Bus).'' Revised Action 34 requires 
that with one degraded grid voltage monitoring channel inoperable, the 
inoperable channel be placed in the tripped condition within one hour; 
otherwise, immediately enter the applicable action statement(s) for the 
associated emergency diesel generator made inoperable by the degraded 
voltage start instrumentation. The revision to Action 34 also requires 
that with two degraded grid voltage monitoring channels inoperable, 
within one hour restore at least one of the channels to operable status 
and place the other channel in the tripped condition; otherwise, the 
associated emergency diesel generator would be declared inoperable and 
its applicable action statement(s) entered. Corresponding changes have 
also been made in the bases for TS 3/4.3.2 and the BVPS-2 TS Index 
pages.
    Date of issuance: May 27, 1998.
    Effective date: Effective immediately, to be implemented within 60 
days (both units).
    Amendment Nos.: 214 and 91.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19969).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 27, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

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

    Date of application for amendments: March 16, 1998, as supplemented 
May 14, 1998.
    Brief description of amendments: These amendments revise technical 
specification (TS) Table 4.3-1 to add footnote 6 to the channel 
calibration requirement for all instrument channels that are provided 
with an input from neutron flux detectors. Footnote 6 provides that 
neutron detectors may be excluded from channel calibrations. In 
addition, BVPS-1 TS Table 4.3-1 is being revised to add channel 
calibration requirements to items 2.b. (Power Range, Neutron Flux, Low 
Setpoint), 5. (Intermediate Range, Neutron Flux), 6. (Source Range, 
Neutron Flux (Below P-10)), and 23. (Reactor Trip System Interlocks P-
6, P-8, P-9, and P-10). Furthermore, changes are being made to correct 
page numbers in the BVPS-2 TS Index and to add corresponding changes to 
the TS Bases for both units.
    Date of issuance: May 28, 1998.
    Effective date: Both units, effective immediately, to be 
implemented within 60 days.
    Amendment Nos.: 215 and 92.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19969).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 28, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

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

    Date of application for amendments: January 9, 1998, as 
supplemented by letter dated April 20, 1998.
    Brief description of amendments: The amendments permit the use of 
fuel with ZIRLO cladding.
    Date of issuance: May 12, 1998.
    Effective date: May 12, 1998.
    Amendment Nos. 196 and 190.
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 25, 1998 (63 
FR 9605).
    The April 20, 1998 letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 12, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

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

    Date of application for amendment: December 10, 1997.
    Brief description of amendment: The amendment clarifies sections of 
the Technical Specifications that have been demonstrated to be unclear 
or conflicting.
    Date of Issuance: June 4, 1998.
    Effective date: June 4, 1998, to be implemented within 30 days.
    Amendment No.: 195.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4313).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated June 4, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station Unit No. 2, Oswego County, New York

    Date of application for amendment: December 15, 1997, as 
supplemented by letter dated April 24, 1998.
    Brief description of amendment: This amendment changes Technical 
Specifications 2.1.2 and 3.4.1.1 to revise the minimum critical power 
ratio safety limits for fuel operating cycle 7 for two-loop and single-
loop recirculation operation.
    Date of issuance: June 4, 1998.
    Effective date: As of the date of issuance to be implemented before

[[Page 33117]]

startup of the Unit 2 reactor to begin fuel operating cycle 7.
    Amendment No.: 82.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4314).
    The April 24, 1998, submittal provided clarifying information that 
did not alter the initial no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 4, 1998.
    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

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: September 2, 1997.
    Brief description of amendment: The amendment corrects several 
compliance issues as identified in Licensee Event Report 97-022-00 
``Technical Specification Violations'' dated July 9, 1997, by rewording 
the text; changing terminology and numbering; combining two Technical 
Specifications (TSs) into one; changing the allowed outage times; 
specifying guidance for entering into TS 3.0.3; changing a definition; 
changing surveillance requirments, and updating the TS Bases section to 
reflect changes.
    Date of issuance: May 26, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 215.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 24, 1997 (62 
FR 50008).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 26, 1998.
    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, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut

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: October 15, 1997, as 
supplemented January 23 and April 8, 1998.
    Brief description of amendment: The amendment revises the action 
statements and the instrumentation trip setpoint tables in the 
Technical Specifications for the reactor trip system and engineered 
safety feature actuation system instrumentation. In addition, the 
amendment (1) decreases the reactor trip setpoint for the reactor 
coolant pump low shaft speed (underspeed trip setpoint) from 95.8 
percent to 92.4 percent of rated speed, (2) makes editorial changes, 
and (3) changes the Bases to reflect the new methodology.
    Date of issuance: May 26, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 159.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 19, 1997 (62 
FR 61842).
    The January 23 and April 8, 1998, submittals provided clarifying 
and additional information that did not change the scope of the October 
15, 1997, application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 26, 1998.
    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, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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: April 7, 1998.
    Brief description of amendment: The amendment replaces the 
pressurizer maximum water inventory requirement with a pressurizer 
maximum indicated level requirement. The amendment also makes editorial 
changes and modifies the associated Bases section.
    Date of issuance: May 27, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 160.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 23, 1998 (63 FR 
20219).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 27, 1998.
    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, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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: April 14, 1998, as supplemented 
May 7, 1998, and two letters dated June 4, 1998.
    Brief description of amendment: The amendment changes Technical 
Specification 3/4.4.4, Relief Valves, to ensure that the automatic 
capability of the power-operated relief valves (PORVs) to relieve 
pressure is maintained when these valves are isolated by closure of the 
block valves. The amendment also makes editorial changes, adds PORV 
surveillance requirements, and modifies the associated Bases section.
    Date of issuance: June 5, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 161.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 20, 1998 (63 FR 
19532).
    The May 7, 1998, letter and the two letters dated June 4, 1998, 
provide clarifying information that did not change the scope of the 
April 14, 1998, application and the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 5, 1998.
    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, and the

[[Page 33118]]

Waterford Library, ATTN: Vince Juliano, 49 Rope Ferry Road, Waterford, 
Connecticut.

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

    Date of application for amendments: December 23, 1997.
    Brief description of amendments: The amendments changed the 
combined Technical Specifications (TS) for the Diablo Canyon Power 
Plant (DCPP) Unit Nos. 1 and 2 to revise TS 3/4.7.1.1, Table 3.7-1, 
``Maximum Allowable Power Range Neutron Flux High Setpoint With 
Inoperable Steam Line Safety Valves.'' The power range (PR) neutron 
flux high setpoints were changed based on revised calculational 
methodologies for 1, 2, or 3 inoperable MSSVs per steam generator (SG). 
The proposed TS change lowered the PR neutron flux high setpoints when 
2 or 3 MSSV are inoperable per loop such that the maximum power level 
allowed would be within the heat removing capability of the remaining 
operable MSSVs. Although the method for calculating the maximum power 
level allowed when one MSSV per loop is inoperable was revised, the 
results were not and the limit remained the same. The associated Bases 
were also revised.
    Date of issuance: May 28, 1998.
    Effective date: May 28, 1998, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 1-125; Unit 2-123.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19975).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 28, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear 
Generating Station, Unit No. 1, Salem County, New Jersey

    Date of application for amendment: March 26, 1998.
    Brief description of amendment: The amendment revises Technical 
Specification 3.1.3.3, ``Rod Drop Time,'' to change the applicability 
from Mode 3 (hot shutdown) to Modes 1 and 2 (startup and power 
operation).
    Date of issuance: June 4, 1998.
    Effective date: As of date of issuance to be implemented within 60 
days.
    Amendment No.: 211.
    Facility Operating License No. DPR-70: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19978). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 4, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

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 30, 1997, as supplemented 
April 1, 1998.
    Brief description of amendments: The amendments revise the 
Technical Specification requirements to reflect a design modification 
that changes the power sources to valves associated with the low 
pressure coolant injection mode of the residual heat removal system.
    Date of issuance: June 2, 1998.
    Effective date: As of the date of issuance to be implemented prior 
to startup from the next refueling outage for both units.
    Amendment Nos.: Unit 1-211; Unit 2-152.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 16, 1997 (62 FR 
38139).
    The April 1, 1998, submittal provided clarifying information that 
did not change the scope of the May 30, 1997, application and the 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia.

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

    Date of application for amendments: November 20, 1997, as 
supplemented by letter dated April 16, 1998.
    Brief description of amendments: The proposed changes to the 
Technical Specifications (TS): (1) Remove the inequalities applied to 
the ``Trip Setpoint'' column of TS Table 3.3.1-1, ``Reactor Trip System 
Instrumentation'' and TS Table 3.3.2-1, ``Engineered Safety Feature 
Actuation System Instrumentation'' and revise the ``Trip Setpoint'' 
column to read ``Nominal Trip Setpoint;'' (2) Add footnotes (n) and (i) 
to TS Tables 3.3.1-1 and 3.3.2-1, respectively, to include criteria for 
channel operability, reset, and calibration tolerance about the trip 
setpoint. These footnotes also allow for the trip setpoint to be set 
more conservatively than the Nominal Trip Setpoint value as necessary 
in response to plant conditions; (3) The Allowable Value for TS Table 
3.3.1-1, Function 14.b, Turbine Trip--Turbine Stop Valve Closure, would 
be revised from ``[greater than or equal to] 96.7% open'' to ``[greater 
than or equal to] 90% open;'' (4) Revise footnotes (l) and (m) of TS 
Table 3.3.1-1 to refer to Nominal Trip Setpoint and delete the 
inequalities applied to the trip setpoints; (5) Delete the superscript 
``(a)'' from the ``Trip Setpoint'' column on page 6 of 8 of Table 
3.3.1-1; (6) Revise the inequality for the Engineered Safety Feature 
Actuation System Allowable Value for Steam Line Pressure--Low (Table 
3.3.2-1, Function 1.e) from ``[less than or equal to]'' to ``[greater 
than or equal to];'' and (7) Revise associated TS Bases to reflect the 
TS revisions.
    Date of issuance: June 1, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1-101; Unit 2-79.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68318).
    The supplement dated April 16, 1998, provided clarifying 
information that did not change the scope of the November 20, 1997, 
application and the initial proposed no significant hazards 
determination.

[[Page 33119]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated June 1, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: May 1, 1995 (TXX-95090).
    Brief description of amendments: These amendments revise section 3/
4.8.1 of the Technical Specifications (TSs) to reduce the minimum fuel 
oil volume requirement during MODES 5 and 6 for an operable emergency 
diesel generator (EDG) and allow continued OPERABLE status of diesel 
generators during all MODES for 48 hours with greater than a 6 day 
supply of diesel fuel for a given EDG.
    Date of issuance: May 22, 1998.
    Effective date: May 22, 1998, to be implemented within 30 days.
    Amendment Nos.: Unit 1--Amendment No. 60; Unit 2--Amendment No. 46.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 21, 1995 (60 FR 
32373).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 22, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

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

    Date of application for amendment: December 4, 1997, as 
supplemented by letters dated January 28, 1998, March 3, 1998, March 9, 
1998, and April 24, 1998.
    Brief description of amendment: The amendment permits the continued 
used of the existing Siemens Power Corporation minimum critical power 
ratio (MCPR) safety limits for WNP-2 Fuel Cycle 14 and changes the ASEA 
Brown Boveri (ABB) MCPR safety limit for single loop operation from 
1.08 for Cycle 13 to 1.09 for Cycle 14.
    Date of issuance: May 29, 1998.
    Effective date: May 29, 1998, to be implemented within 30 days from 
the date of issuance.
    Amendment No.: 154.
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 14, 1998 (63 FR 
2284).
    The January 28, 1998, March 3, 1998, March 9, 1998, and April 24, 
1998, 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 May 29, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352.

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

    Date of application for amendment: February 25, 1998.
    Brief description of amendment: The amendment revises the Technical 
Specifications to implement performance-based containment leakage 
testing under Option B of 10 CFR 50, Appendix J.
    Date of issuance: May 28, 1998.
    Effective date: May 28, 1998.
    Amendment No.: 136.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 8, 1998 (63 FR 
17237).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 28, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.

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

    Date of amendment request: May 8, 1998, as supplemented by letter 
dated May 11, 1998.
    Brief description of amendment: The amendment adds a new Action 
Statement to Technical Specification 3/4.3.2, Table 3.3-3, Functional 
Unit 7.b., Refueling Water Storage Tank Level--Low-Low Coincident With 
Safety Injection.
    Date of issuance: May 28, 1998.
    Effective date: May 28, 1998.
    Amendment No.: 117.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (63 FR 26829 dated May 14, 1998). 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 June 15, 1998, 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, consultation with the State of Kansas and final 
determination of no significant hazards consideration are contained in 
a Safety Evaluation dated May 28, 1998.
    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.

    Dated at Rockville, Maryland, this 10th day of June 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-16012 Filed 6-16-98; 8:45 am]
BILLING CODE 7590-01-P