[Federal Register Volume 64, Number 183 (Wednesday, September 22, 1999)]
[Notices]
[Pages 51343-51356]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-24573]


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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 August 28, 1999, through September 10, 1999. 
The last biweekly notice was published on September 8, 1999 (64 FR 
48858).

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)

[[Page 51344]]

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 October 22, 1999, 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, 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

[[Page 51345]]

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.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of amendments request: August 27, 1999.
    Description of amendments request: The proposed amendment revises 
Technical Specification (TS) 3.7.13, ``Spent Fuel Pool (SFP) Water 
Level'' to allow placement of one or more fuel assemblies on SFP rack 
spacers to support fuel reconstitution activities while irradiated fuel 
assembly movement continues in the SFP. Although the plant TSs do not 
prohibit fuel reconstitution, the effect of the current wording of TS 
3.7.13, in conjunction with the specific design of the SFP and storage 
racks, limits reconstituting only one fuel assembly at a time and only 
when no irradiated fuel assembly movement occurs in the SFP. 
Specifically, the proposed change adds a new statement to the limiting 
condition for operation that would require the water level over fuel 
assemblies placed on rack spacers to be 19.8 feet while irradiated fuel 
assemblies are being moved in the SFP. The proposed administrative 
controls will ensure that the current design basis fuel handling 
accident described in the Updated Final Safety Analysis Report (UFSAR) 
bounds a fuel handling accident associated with reconstitution 
activities.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed change will require a minimum water level of 19.8 
feet over fuel assemblies that are placed on rack spacers for fuel 
reconstitution activities while fuel movement continues in the SFP. 
This proposed change does not cause any spent fuel handling 
equipment to be operated in a new or different manner. No structural 
changes or modifications are being made to the spent fuel handling 
machine (SFHM) or to the spent fuel storage racks. Administrative 
controls will be put in place to ensure that the SFHM or an assembly 
being carried by the SFHM will not strike assemblies placed on rack 
spacers. This proposed change does not make any changes to 
equipment, procedures, or processes that increase the likelihood of 
dropping the fuel assembly from the SFHM. Administrative controls 
will be put in place to limit the movement of heavy loads such that 
only a single-failure-proof crane will be used in the area of the 
affected fuel assembly and the adjacent storage rack cells when the 
assemblies are seated on rack spacers with their upper end fittings 
removed. Therefore, this proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    A Fuel Handling Incident (FHI) during reconstitution activities 
is bounded by those previously analyzed and described in the Updated 
Final Safety Analysis Report (UFSAR) for the limiting FHI. The 
number of fuel pins that could be ruptured in a raised fuel assembly 
does not exceed that previously analyzed. Also, by requiring that 
reconstitution activities do not occur until 10 days after shutdown 
ensures that a[n] FHI during these activities will be bounded by the 
most limiting FHI described in the UFSAR. Therefore, the proposed 
change does not significantly increase the consequences of an 
accident previously evaluated.
    Based on the above, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The proposed change will not make any physical changes to the 
plant. Specifically, no modifications will be made to the SFHM, the 
spent fuel storage racks, or the spent fuel assemblies. No changes 
are made to the operation of the SFHM. The only change made by this 
activity is that multiple fuel assemblies may be placed on rack 
spacers in the SFP for reconstitution activities. Administrative 
controls will be put in place to ensure that this proposed change 
does not create the potential of a[n] FHI during reconstitution 
activities that is not bounded by our current accident analysis. 
This proposed change does not have any impact on the cooling or safe 
geometry functions of the SFP storage racks. This proposed change 
does not create any new interactions between any plant components. 
Therefore, the possibility of a new or different type of accident is 
not created by this proposed change.
    3. Would not involve a significant reduction in a margin of 
safety.
    The Technical Specification requires a minimum water level to be 
maintained above the fuel assemblies stored in the SFP storage racks 
to ensure that sufficient water depth is available to remove the 
assembled iodine gap activity released from the rupture of an 
irradiated fuel assembly. The proposed change will allow multiple 
fuel assemblies to be placed on rack spacers for fuel reconstitution 
activities while fuel movement continues in the spent fuel pool. 
These activities will reduce the amount of water maintained above 
the fuel assemblies that are placed on rack spacers. However, the 
proposed change does not involve a significant reduction in a margin 
of safety based on the administrative controls that require an 
increase in the decay time before these activities can be started. 
Additional administrative controls will be put in place that 
include, in part, restricting load movements over the affected fuel 
assembly and the adjacent storage rack cells, as well as controlling 
the SFHM. The administrative controls will ensure that the FHI 
associated with reconstitution activities is bounded by the current 
design basis FHI described in the UFSAR. Therefore, the proposed 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for Licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: S. Singh Bajwa.

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

    Date of amendment request: August 18, 1999.
    Description of amendment request: The proposed amendment will 
change the required surveillance interval for cycling the steam valves 
in the turbine overspeed protection system from monthly to quarterly. 
The license requirement is documented in the St. Lucie, Unit 2 Updated 
Final Safety Analysis Report (UFSAR) Section 13.7.1.6.2, and the 
proposed change does not satisfy the 10 CFR 50.59 standards for a 
change that can be made by the licensee without prior Commission 
approval.
    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 small increase in turbine missile ejection frequency 
resulting from extending the test interval for turbine valves is 
acceptable with respect to the NRC probabilistic acceptance 
criterion and supports quarterly testing. In addition, there are no 
physical changes to plant equipment or changes in plant operation 
that could initiate or adversely affect the mitigation or

[[Page 51346]]

consequences of an accident previously evaluated. Turbine disk 
integrity remains unchanged since the turbine rotor inspection cycle 
is not affected by the change in valve testing frequency. Further, 
there are no changes to protective barriers or changes in separation 
of equipment important to safety. Therefore, safety related 
structures, systems, and components remain adequately protected 
against potential turbine missiles and the potential for turbine 
missile generation has not significantly increased. The change to 
extend the turbine valve test interval maintains the intent and 
design basis function being verified by the surveillance 
requirement. Therefore, operation of the facility in accordance with 
the proposed amendment will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    There are no physical changes to plant equipment or changes in 
plant operation that could create a new or different kind of 
accident. This proposed change does not result in any plant 
configuration changes or create new failure modes. The small 
increase in turbine missile ejection frequency resulting from 
extending the test interval for turbine valves is acceptable with 
respect to the NRC probabilistic acceptance criterion and supports 
quarterly testing. New types of turbine missiles or strike 
probabilities are not created by extending the turbine valve test 
interval. No new or different kind of accident is created. In 
addition, turbine disk integrity remains unchanged since the turbine 
rotor inspection cycle is not affected by the change in valve 
testing frequency. Further, there are no changes to protective 
barriers or changes in the separation of equipment important to 
safety. Safety related structures, systems, and components remain 
adequately protected against potential turbine missiles, the 
potential for turbine missile generation has not significantly 
increased, and new or different kinds of accidents are not created. 
Therefore, operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    This proposed surveillance change extends the turbine overspeed 
protection system turbine valve test frequency from monthly to 
quarterly. The results of turbine missile ejection frequency remain 
within NRC acceptance criterion and therefore supports quarterly 
testing. There are no physical changes to plant equipment or changes 
in plant operation that involve a significant reduction in the 
margin of safety. Turbine disk integrity remains unchanged since the 
turbine rotor inspection cycle is not affected by the change in 
valve testing frequency. There are no changes to protective barriers 
or changes in separation of equipment important to safety. 
Therefore, safety related structures, systems, and components remain 
adequately protected against potential turbine missiles and the 
potential for turbine missile generation has not significantly 
increased. The change in turbine valve test interval maintains the 
intent and design basis function being verified by the surveillance 
requirement. As such, the assumptions and conclusions of the 
accident analyses in the UFSAR remain valid and associated safety 
limits will continue to be met. Therefore, operation of the facility 
in accordance with the proposed amendment would not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Sheri R. Peterson.

GPU Nuclear Inc., Docket No. 50-320, Three Mile Island--Unit 2 (TMI-2), 
Dauphin County, Pennsylvania

    Date of amendment request: June 29, 1999, as supplemented August 
27, 1999 (LAR No. 77).
    Description of amendment request: The proposed amendment would 
grant authority for the licensee to possess limited amounts and types 
of radioactive materials without unit distinction so that after the 
sale and transfer of the Three Mile Island--Unit 1 (TMI-1) license to 
AmerGen, radioactive materials may continue to be moved between the 
TMI-1 and TMI-2 units. After the license transfer, GPU Nuclear will 
need to access the waste handling and processing facilities at TMI-1 
(currently common facilities) for its normal post-defueling monitored 
storage (PDMS) activities. Similarly, AmerGen as the TMI-1 licensee and 
PDMS contractor, will need to move radioactive apparatus and materials 
between units, principally during TMI-1 outages. The amendment would 
not authorize receipt or possession of radioactive material or waste 
from other sites.
    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 would not involve a significant increase 
in the probability of an accident previously evaluated because no 
accident initiators or assumptions are affected. The proposed 
changes have no effect on any plant systems. All Limiting Conditions 
for PDMS and Safety Limits specified in the Technical Specifications 
will remain unchanged.
    [The proposed changes would] not involve a significant increase 
in the consequences of an accident previously evaluated because no 
accident conditions or assumptions are affected. The proposed 
changes do not alter the source term, containment isolation, or 
allowable radiological consequences. The staging of radioactive 
materials such as the contaminated reactor coolant pump and motor 
components will not result in a source term, that if released, would 
exceed that previously analyzed in the PDMS SAR [safety analysis 
report] in terms of off-site dose consequences. The proposed changes 
have no adverse effect on any plant system.
    2. [The proposed changes would] not create the possibility of a 
new or different kind of accident from any previously evaluated 
because no new accident initiators or assumptions are introduced by 
the proposed changes. The proposed changes have no direct effect on 
any plant system. The changes do not affect any system functional 
requirements, plant maintenance, or operability requirements.
    3. [The proposed changes would] not involve a significant 
reduction in the margin of safety because the proposed changes do 
not involve significant changes to the initial conditions 
contributing to accident severity or consequences. The proposed 
changes have no direct effect on any plant systems.

    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: Law/Government Publications 
Section, State Library of Pennsylvania, (Regional Depository) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Michael T. Masnik.

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit 
1, DeWitt County, Illinois

    Date of amendment request: August 23, 1999.
    Description of amendment request: The proposed amendment would 
delete certain license conditions that are obsolete and no longer 
apply.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 51347]]

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 activity does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    The proposed changes delete various license conditions each of 
which has been fulfilled and no longer warrants a license condition. 
As such, the changes are purely administrative in nature, and 
involve no physical or operational changes to the facility. The 
initial conditions and methodologies used in the accident analyses 
consequently remain unchanged. Further, the proposed changes do not 
change or alter the design assumptions for the systems or components 
used to mitigate the consequences of an accident. Therefore, 
accident analyses results are not impacted. On this basis, the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) The proposed activity does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    As noted above, the proposed changes are purely administrative 
and involve no physical or operational changes to the facility. As 
such, the proposed changes do not affect the design or operation of 
any system, structure, or component in the plant. The safety 
functions of the related structures, systems, or components are not 
changed in any manner, nor is the reliability or[f] any structures, 
systems or components reduced. No new or different type of equipment 
will be installed, and consequently, no new failure modes are 
introduced. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) The proposed activity does not involve a significant 
reduction in the margin of safety.
    The proposed changes are administrative in nature and have no 
impact on the margin of safety of any Technical Specification. There 
is no impact on safety limits or limiting safety system settings. 
The changes do not affect any plant safety parameters or setpoints. 
All active/applicable license conditions set forth in the CPS 
Operating License will remain in effect, and no physical or 
operational changes to the facility will result from these changes. 
Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, IL 61727.
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, IL 62525.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: August 26, 1999.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to reflect the proposed 
implementation of Noble Metal Chemical Addition (NMCA) so as to enhance 
the effectiveness of Hydrogen Water Chemistry (HWC) in mitigating 
Intergranular Stress Corrosion Cracking (IGSCC) in reactor vessel 
internal components. Specifically, the proposed amendment would raise 
the reactor water conductivity limit in TS 3.2.3.a from 1.0 micromho/cm 
to 20 micromho/cm and in TS 3.2.3.c.1 from 5.0 micromho/cm to 20.0 
micromho/cm during NMCA application. The proposed amendment will also 
raise the limit in TS 3.2.3.a and 3.2.3.b from 1 micromho/cm to 2 
micromho/cm for up to a 5-month period at power operation following 
NMCA application. The reactor water conductivity would be restored to 
within the limit currently specified in TS 3.2.3 after the NMCA process 
is complete. The Bases for TS 3.2.3 and 4.2.3, ``Coolant Chemistry,'' 
would be supplemented to explain the changes resulting from NMCA.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment to TS 3.2.3 will raise the reactor water 
conductivity limit during and following NMCA application. This 
change will allow the application of a layer of noble metals to the 
reactor vessel internals to enhance the effectiveness of HWC in 
mitigating IGSCC. An increased conductivity is expected both during 
and following NMCA. However, during NMCA, this increase is caused 
principally by residual ionic species which do not contribute to 
IGSCC. Following NMCA application, the increased conductivity is 
expected to be due to soluble iron and increased pH which has no 
adverse affect on crack growth. Accordingly, the proposed change 
will not adversely affect reactor vessel internals or reactor fuel 
such that the probability of an accident is increased. The proposed 
change will not alter the current TS requirements concerning 
equipment needed to mitigate the consequences of an accident nor 
affect the performance of this equipment. Therefore, operation in 
accordance with the proposed amendment will not create an increase 
in the probability or consequences of an accident previously 
evaluated.
    The operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed amendment to TS 3.2.3 will raise the reactor water 
conductivity limit during and following NMCA application. This 
change will allow the application of a layer of noble metals to the 
reactor vessel internals to enhance the effectiveness of HWC in 
mitigating IGSCC. Except for these temporary exceptions to the 
existing reactor coolant chemistry specification, no new plant or 
system operating modes are being introduced and plant equipment will 
continue to perform their intended function. An increased 
conductivity is expected both during and following NMCA. However, 
during NMCA, this increase is caused by ionic species which do not 
contribute to IGSCC. Following NMCA application, the increased 
conductivity is due to soluble iron and increased pH which has no 
adverse affect on crack growth. Accordingly, the proposed changes 
will not affect plant equipment in a way to create a new or 
different kind of accident. Therefore, operation 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 operation of Nine Mile Point Unit 1, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The proposed amendment to TS 3.2.3 will raise the reactor water 
conductivity limit during and following the application of NMCA. 
During NMCA, the proposed change will raise the reactor water 
conductivity limit in TS 3.2.3a and 3.2.3c.1 to 20 [micromho/cm]. 
However, the expected increase in coolant conductivity is caused 
principally by ionic species which do not contribute to IGSCC and, 
therefore, will not adversely affect reactor vessel internals or 
reactor fuel.
    Following NMCA application, industry experience indicates that 
there may be an elevated conductivity approaching the 1 [micromho/
cm] conductivity limit delineated in TS 3.2.3a and 3.2.3b. To 
provide operating margin, NMPC proposes to raise this limit to 2 
[micromho/cm] for up to 5 months of power operation following 
application. The expected increase in the conductivity is attributed 
to an increase in soluble iron and pH in the reactor coolant which 
results from the application of the noble metals and its affect on 
the deposits on the fuel. Soluble iron nor increased pH contribute 
to IGSCC crack growth. The existing 1 [micromho/cm] limit is based 
on EPRI [Electric Power Research Institute] guidelines action Level 
2 for power operation, which assumes normal

[[Page 51348]]

conductivity below .3 [micromho/cm]. Increasing the limit to 2 
[micromho/cm] during the period when soluble iron levels are high 
provides an equivalent operating margin consistent with the chloride 
and sulfate limits. Accordingly, this temporary ([less than] 5 
months) elevated conductivity is expected, acceptable, and not 
considered ``abnormal'' as discussed in TS 4.2.3 and associated 
Bases. Daily samples of coolant for conductivity, chlorides and 
sulfates will continue to be performed to assure water quality.
    Therefore, operation in accordance with the proposed amendment 
will not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: S. Singh Bajwa.

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

    Date of amendment request: August 26, 1999
    Description of amendment request: The proposed amendment would 
raise the condensate storage tank (CST) low level setpoint and the 
corresponding allowable value in Technical Specification (TS) Tables 
3.3.3-2 and 3.3.5-2. The subject setpoint is associated with the 
automatic transfer of the High Pressure Coolant Injection (HPCI) and 
Reactor Core Isolation Cooling (RCIC) pump suctions from the CST to the 
suppression pool in the event of low CST level. These changes are being 
made to address concerns regarding potential vortexing in the HPCI and 
RCIC suction flowpaths.
    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 systems affected by the proposed change provide accident 
mitigation functions. Neither the proposed increase in level 
setpoint nor the reliance on operator action to maintain the 
required 135,000 gallon reserve volume in the condensate storage 
tank (CST) can affect initiation of a design basis accident.
    Raising the CST low level setpoint to account for potential 
vortexing in the HPCI and RCIC suction flowpaths provides assurance 
that the functions of these systems can be properly carried out. 
There will no longer be a possibility of air entrainment into the 
RCIC and HPCI pumps suction at low levels in the CST. Initiation of 
RCIC or HPCI flow is unaffected by this modification. Execution of 
the suction line transfer to the suppression pool remains an 
entirely automatic function, utilizing the same safety related 
instrument signals as previously.
    Reliance on level alarms and operator action to maintain the 
135,000-gallon minimum reserve water volume in the CST, in lieu of 
internal standpipes, cannot increase the consequences of an 
accident. This is an operational condition that establishes initial 
conditions prior to an accident occurring. Operators would have 
sufficient time to respond to a CST level decrease under non-
accident conditions. Manually transferring HPCI and RCIC suction to 
the safety related suppression pool should CST level decline below 
203,000 gallons (the 135,000 gallons required inventory, plus 68,000 
gallons unusable) ensures HPCI and RCIC remain fully capable of 
performing their design basis functions.
    All parameters pertaining to the accident analysis, including 
pump initiation time, flowrate, volume and duration of flow 
delivered to the reactor vessel remain satisfied following 
implementation of this proposed change. Therefore, no accident 
scenario evaluated in the SAR [Safety Analysis Report] will be 
affected, and the radiological consequences of accidents previously 
evaluated in the SAR are not increased.
    These changes, therefore, do not modify or add any initiating 
parameters that would significantly increase the probability or 
consequences of any previously analyzed accident.
    (2) The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Implementation of these proposed changes cannot create the 
possibility of a different type of accident from any previously 
considered. First, the affected systems only perform mitigation 
functions, so postulated failures of any of these systems would not 
initiate a design basis accident. The function credited in the 
safety analysis is automatic transfer of the HPCI and RCIC suction 
lines from the CST to the suppression pool. This automatic transfer 
will still occur as required, with the only difference being 
execution earlier at a higher CST water level. Any considerations 
associated with maintaining the required minimum CST water level, 
including reliance on an alarm and operator action in lieu of a 
passive design feature, cannot lead to an accident of a different 
type since the CST itself is explicitly excluded from consideration 
in the accident analysis. Although the preference is to provide 
shutdown cooling with the reactor grade water of the CST, failure to 
do so will neither impact the ability to achieve shutdown cooling 
nor create a new type of accident.
    Therefore, the proposed changes do 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.
    The margin of safety of the affected TS is maintained. RCIC is 
provided to assure adequate core cooling in the event of reactor 
isolation from its primary heat sink and concurrent loss of 
feedwater flow to the reactor vessel without requiring actuation of 
ECCS [Emergency Core Cooling System] equipment. This function will 
be accomplished. HPCI provides a backup to RCIC for safe shutdown 
and the ECCS function of ensuring the reactor core is adequately 
cooled to limit fuel clad temperature during a small break loss of 
coolant accident. The safety analysis does not credit CST water. 
Since the automatic transfer to the suppression pool is assured with 
the same high quality and reliability as before, the ECCS function 
is not affected. Should CST level decline below the required minimum 
volume, operators would align HPCI and RCIC suction to the 
suppression pool. System design functions, including containment 
isolation, continue to be maintained in this alignment.
    The CST also provides a source of water for shutdown during 
station blackout (SBO) scenarios. The proposed changes do not affect 
the ability to recover from a SBO scenario.
    Core spray is provided to assure that the core is adequately 
cooled following a LOCA [Loss of Coolant Accident] and provides core 
cooling capacity for all break sizes. Core spray is a primary 
cooling source after the reactor vessel is depressurized and a 
source for flooding in case of accidental draining. In Operational 
Conditions 4 or 5, the CST is relied upon as the cooling water 
source if the suppression pool is drained below its minimum level. 
Operator actions in response to a CST alarm ensure sufficient 
condensate inventory is available to accomplish this function.
    ECCS instrumentation (HPCI) is provided to initiate actions to 
mitigate the consequences of accidents that are beyond the ability 
of the operator to control. RCIC instrumentation is provided to 
initiate actions to assure adequate core cooling in the event of 
reactor isolation from its primary heat sink and the loss of 
feedwater flow to the reactor vessel. The HPCI and RCIC level 
instruments continue to provide their automatic function thereby 
preserving the design requirements of these systems. Remote shutdown 
instrumentation and controls ensure that sufficient capability is 
available to permit shutdown and maintenance of Hot Shutdown of the 
unit from locations outside the control room in the event control 
room habitability is lost. RCIC continues to satisfy this function.
    All design basis requirements of HPCI, RCIC, core spray and the 
CST continue to be satisfied to ensure safe shutdown and

[[Page 51349]]

mitigate a LOCA. Required water volumes remain available for core 
cooling, as is the automatic transfer to the safety related 
suppression pool source.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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 Section Chief: James W. Clifford.

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

    Date of amendment request: July 29, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Surveillance Requirement 4.6.1.1 to 
clarify when verification of primary containment integrity may be 
performed by administrative means and to change the surveillance 
interval for verification of manual valves and blind flanges inside of 
containment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. The operation of Salem Nuclear Generating Station, Unit Nos. 
1 and 2, in accordance with the proposed amendment will not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    The licensee has determined that the proposed change will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated. The proposed change revises means 
for verification of containment integrity in certain cases by 
allowing the verification to be conducted by administrative means 
such as tagging requests, other TS surveillance procedures and 
previously performed valve alignments. Although the current Salem 
TSs allow the use of administrative means to verify valve position, 
its application is limited to valves that are open under 
administrative controls.
    The proposed amendment does not change the position of 
containment isolation valves or otherwise modify the containment 
integrity. Thus, the assumptions made in evaluating the occurrence 
and radiological consequences of accidents described in the Safety 
Analysis Report (SAR) have not been changed. The proposed change to 
use administrative means continues to ensure that the release of 
radioactive materials from the containment atmosphere will be 
restricted to those leakage paths and associated leak rates assumed 
in the accident analysis. Allowing the use of administrative means 
to verify compliance with the surveillance requirement for these 
valves is acceptable based on the limited access to these areas in 
Modes 1 through 4 (power operation through hot shutdown). The 
probability of misalignment of these containment isolation valves, 
once they have been verified in the proper position is small. The 
probability of occurrence of any previously evaluated accident is 
independent of valve position verification.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated in the SAR.
    2. The operation of Salem Nuclear Generating Station, Unit Nos. 
1 and 2, in accordance with the proposed amendment does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    The licensee has determined that the proposed amendment does not 
physically alter the facility or change the operation of the 
facility. The proposed change does not affect the current operation 
and response of any systems, structures or components assumed to 
function in the accident analysis. Additionally, the proposed change 
does not increase the consequences of a malfunction of equipment 
important to safety. The proposed change to use administrative means 
in lieu of field verification continues to ensure that the release 
of radioactive materials from the containment atmosphere will be 
restricted to those leakage paths and associated leak rates assumed 
in the accident analysis.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The operation of Salem Nuclear Generating Station, Unit Nos. 
1 and 2, in accordance with the proposed amendment does not involve 
a significant reduction in a margin of safety.
    The licensee has determined that the proposed amendment does not 
involve a significant reduction in a margin of safety. The proposed 
change involves a revision of certain TSs surveillance requirements 
and frequency of performance. The proposed change does not modify 
hardware or plant operation, and the accident analyses are 
unchanged. The proposed amendment will continue to ensure that the 
proper valves are identified and tested in accordance with the TS 
requirements. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.

    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038
    NRC Section Chief: James W. Clifford

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

    Date of amendment request: August 25, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Appendix C, ``Additional 
Conditions,'' to authorize the performance of single cell charging of 
operable safety-related batteries by using non-Class 1E single cell 
battery chargers, with proper electrical isolation. The single cell 
chargers would be used to restore individual cell float voltage to the 
normal TS limit.
    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 change permits the use of an industry accepted 
method to restore a battery cell to its design basis from an 
OPERABLE but degraded condition or to prevent a cell from becoming 
degraded. IEEE Std [Institute of Electrical and Electronics 
Engineers Standard] 450-1995, ``IEEE Recommended Practice for 
Maintenance, Testing, and Replacement of Vented Lead Storage 
Batteries for Stationary Applications,'' states that single cell 
charging is an acceptable method of correcting low cell voltage or 
low specific gravity conditions for a single cell or for a small 
number of cells.
    At least two class 1E fuses in series will be used on both the 
positive and negative leads between the battery and the charger to 
protect the battery if a fault should develop in the charger. The 
battery charger design includes diodes, a power transformer and 
control circuitry to prevent draining the connected cells in the 
event of a short circuit in the 120 Volt ac source or a loss of 
charger input or output voltage. Charger output is controlled 
automatically to prevent overcharging the connected cells.
    In the event of a controller failure resulting in charger 
overvoltage, procedural controls

[[Page 51350]]

governing the use of the charger ensure the condition is detected 
and corrected before failure of a connected cell occurs. While the 
single cell charger is connected, procedures will require periodic 
checks to verify proper charger operation and to measure electrolyte 
level, temperature and specific gravity for the cells being charged. 
Monitoring will be performed at least once every eight hours, a 
frequency sufficient to ensure compliance with the requirements of 
the Technical Specifications.
    An insulating material will be used to minimize the possibility 
of shorting leads or clips at the battery. Administrative controls 
governing the use and storage of transient loads are sufficient to 
ensure the use of single cell battery chargers does not create a 
potential missile hazard to safety related systems, structures and 
components.
    The Class 1E DC system is not an accident initiator. The Class 
1E DC system supports the operation of safety related equipment 
required for the safe shutdown of the plant and for the mitigation 
of accident conditions. Therefore, the proposed change does not 
increase the probability of an accident previously evaluated.
    The station's dc systems will be operable to mitigate the 
consequences of an accident previously evaluated. Single cell 
charging would be limited to one OPERABLE class 1E battery bank at a 
time for either the 28 VDC or 125 VDC systems. Therefore, failure of 
a class 1E battery as a result of single cell charging would be 
limited to a single channel and would not reduce the number of 
OPERABLE dc sources below that required to safely shutdown the 
plant. Administrative controls would also prohibit the use of single 
cell charging for an OPERABLE class 1E battery if less than the 
minimum number of class 1E batteries required by Technical 
Specifications are OPERABLE.
    The proposed change does not cause the capability of the class 
1E DC system to be degraded below the level assumed for any accident 
described in the SAR [Safety Analysis Report]. It would enhance the 
availability of safety related equipment required for the safe 
shutdown of the plant and for the mitigation of accident conditions. 
Therefore the radiological consequences of an accident will remain 
inside the design basis while single cell charging is performed on 
an OPERABLE battery.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The potential to adversely affect the Class 1E batteries is 
minimized by the use of Class 1E fuses and by appropriate 
administrative controls. Failure modes associated with the proposed 
change are bounded by the loss of a Class 1E battery bank which was 
previously evaluated. Therefore, the proposed change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change permits the use of non-Class 1E single cell 
battery chargers, with proper electrical isolation, for charging 
connected cells in OPERABLE class 1E batteries. This would allow 
parameters for an individual cell or for a small number of cells to 
be restored to the normal values specified in Technical 
Specifications without affecting the remainder of the cells in the 
battery. Increased cell monitoring after single cell charging, 
together with PSE&G's corrective action program which requires 
degraded and non-conforming conditions to be documented and 
evaluated, provides assurance that the use of single cell charging 
will not cause long-term cell degradation to go undetected. Since 
all battery cells are required to be maintained within the allowable 
values specified in Technical Specifications, and since the use of 
the single cell charger will not adversely affect battery capacity 
or capability, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

Previously Published Notice 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.

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

    Date of amendment request: July 30, 1999.
    Description of amendment request: The proposed amendments would 
temporarily change the Technical Specifications (TS) to increase the 
upper temperature limit for the Ultimate Heat Sink (UHS) from 98 
degrees Fahrenheit to 100 degrees Fahrenheit. The proposed temporary 
change would be in effect until September 30, 1999.
    Date of publication of individual notice in Federal Register: 
August 18, 1999 (64 FR 44962).
    Expiration date of individual notice: September 17, 1999.
    Local Public Document Room location: Wilmington Public Library, 201 
S. Kankakee Street, Wilmington, Illinois 60481.

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.

[[Page 51351]]

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County, 
Maryland

    Date of application for amendments: November 30, 1998, as 
supplemented May 25, 1999.
    Brief description of amendments: The amendments revise the 
appropriate Technical Specifications to permit the use of leak-limiting 
Alloy 800 repair sleeves developed by AAB--Combustion Engineering (ABB-
CE) to be used at Calvert Cliffs.
    Date of issuance: September 1, 1999.
    Effective date: As of the date of issuance to be implemented during 
the spring 2000.
    Amendment Nos.: 231 and 207.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 13, 1999 (64 FR 
2244).
    The May 25, 1999, letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated September 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendment: January 28, 1999.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 5.6.5, ``Core Operating Limits Report (COLR),'' to 
add two references to the list of approved topical reports.
    Date of issuance: September 1, 1999.
    Effective date: September 1, 1999.
    Amendment No.: 185.
    Facility Operating License No. DPR-23. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9184).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 1, 1999.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

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

    Date of application for amendment: January 22, 1999.
    Brief description of amendment: The amendment revises Technical 
Specifications 4.3.a and 4.3.b and Basis Section 4.3 to permit reactor 
coolant system leak test to be performed at normal operating pressure 
following each refueling outage according to the requirement of the 
American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code, Section XI, and implemented in accordance with 10 CFR 50.55a(g).
    Date of issuance: September 2, 1999.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 203.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 7, 1999 (64 FR 
17023).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 2, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: June 17, 1998, as supplemented 
June 23 and December 2, 1998, and March 18, 1999.
    Brief description of amendment: The amendment revises the Technical 
Specifications to reduce the minimum reactor vessel flow rate 
requirement and revise the units of measurement for consistency with 
the flow measurement procedure.
    Date of issuance: September 3, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No.: 187.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 2, 1998 (63 FR 
36271).
    The December 2, 1998, letter provided additional clarifying 
information and the March 18, 1999, letter requested a 60-day allowance 
for implementation of the amendment. The additional information and 
proposed change to the implementation period were within the scope of 
the original Federal Register notice and did not change the staff's 
initial proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423-3698.

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: December 24, 1998, as 
supplemented June 15, June 17, and July 7, 1999.
    Brief description of amendments: The amendments revise the 
Technical Specification (TS) requirements for the axial flux difference 
(AFD) monitor, quadrant power tilt ratio (QPTR) monitor, rod position 
deviation monitor, and rod insertion limit (RIL) monitor. Specifically, 
the changes (1) relocate requirements for the AFD monitor and the QPTR 
monitor to the Licensing Requirements Manual; (2) delete requirements 
for the rod position deviation monitor and RIL monitor from the TSs; 
(3) modify Unit 1 surveillance requirements (SR) 4.1.3.5 and 4.1.3.6 by 
incorporating the Unit 2 wording to provide surveillances more 
consistent with the Limiting Condition for Operation; (4) change Unit 1 
SR 4.1.3.2.2, SR 4.1.3.5, SR 4.1.3.6 and Unit 2 SR 4.1.3.5 from 24-hour 
surveillance frequencies to 12-hour frequencies; and (5) delete Unit 1 
SR 4.1.3.2.3.
    Date of issuance: August 30, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 225 and 102.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 27, 1999 (64 FR 
4155) The June 15, June 17, and July 7, 1999, letters provided 
additional information but did not change the initial proposed no 
significant hazards consideration determination or expand the amendment 
beyond the scope of the initial notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 30, 1999.
    No significant hazards consideration comments received: No

[[Page 51352]]

    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

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

    Date of amendment request: April 9, 1999, as supplemented by letter 
dated July 14, 1999.
    Brief description of amendment: Revises requirements affecting the 
surveillance methods for the containment tendons, the conduct of 
containment visual inspections, and the reporting methods employed in 
disseminating the results of these inspections to the NRC.
    Date of issuance: September 9, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment No.: 199.
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 19, 1999 (64 FR 
27320).
    The July 14, 1999, letter provided clarifying information that did 
not change the scope of the April 9, 1999, 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 September 9, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit 1, Lake County, Ohio

    Date of application for amendment: March 17, 1999.
    Brief description of amendment: This amendment approves a proposed 
modification that changes the Perry facility as described in the 
Updated Safety Analysis Report. The change incorporates a leak-off line 
in the residual heat removal system. The leak-off line is designed to 
eliminate an operator work around, which will significantly reduce the 
collective dose to operations personnel.
    Date of issuance: August 31, 1999.
    Effective date: August 31, 1999.
    Amendment No.: 106.
    Facility Operating License No. NPF-58: This amendment authorizes 
the revision of the Updated Safety Analysis Report.
    Date of initial notice in Federal Register: May 19, 1999 (64 FR 
27322)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 31, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, OH 44081

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

    Date of application for amendment: May 10, 1999.
    Brief description of amendment: The amendment corrects an invalid 
reference in Section 5.8, ``High Radiation Area,'' of the Crystal River 
Unit 3 Improved Technical Specifications (ITS).
    Date of issuance: September 3, 1999.
    Effective date: September 3, 1999.
    Amendment No.: 186.
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38026)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.

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: May 17, 1999.
    Brief description of amendment: The amendment revises Technical 
Specification (TS) section 4.4.6.2.2.e to replace the reference to 
American Society of Mechanical Engineers (ASME) Code paragraph IWV-
3472(b) which pertains to the frequency of leakage rate testing for 6-
inch, nominal pipe size valves and larger with the requirement that the 
surveillance interval and frequency of surveillance leakage rate 
testing for these valves be performed pursuant to the requirements of 
TS 4.0.5, ``Operations and Surveillance Requirements.''
    Date of issuance: September 10, 1999.
    Effective date: As of the date of issuance.
    Amendment No.: 174.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38033).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 10, 1999.
    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.

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

    Date of application for amendments: May 13, 1999.
    Brief description of amendments: The amendments revise Technical 
Specifications 6.2.A.2, ``Onsite and Offsite Organizations,'' to 
reflect a change in the plant organizational structure that was 
implemented on March 1, 1999.
    Date of issuance: August 26, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 146 and 137.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38034).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

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

    Date of application for amendment: April 12, 1999.
    Brief description of amendment: The amendment removes from the 
Technical Specifications a footnote regarding departure from nucleate 
boiling analysis.
    Date of issuance: September 2, 1999.
    Effective date: September 2, 1999.

[[Page 51353]]

    Amendment No.: 191.
    Facility Operating License No. DPR-64: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 19, 1999 (64 FR 
27324).
    No significant hazards consideration comments received: No.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 2, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: January 28, 1999, as 
supplemented May 4, 1999
    Brief description of amendment: The amendment changes the reactor 
trip on turbine trip from at or above 10 percent rated power to at or 
above the P-8 setpoint.
    Date of issuance: September 8, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 192.
    Facility Operating License No. DPR-64: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 21, 1999 (64 FR 
19563).
    The May 4, 1999, letter provided additional information that did 
not change the staff's proposed finding of no significant hazards 
consideration.
    No significant hazards consideration comments received: No.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: March 29, 1999, as supplemented 
June 21, 1999.
    Brief description of amendment: This amendment revises the 
Technical Specifications (TSs) by relocating the procedural details of 
the Radiological Effluent Technical Specifications (RETS) to the 
Offsite Dose Calculation Manual. The TSs were also revised to relocate 
procedural details associated with solid radioactive wastes to the 
Process Control Program. In addition, the Administrative Controls 
section of the TSs was revised to incorporate programmatic controls for 
radioactive effluents and environmental monitoring.
    These changes are consistent with the guidance provided in Generic 
Letter 89-01, ``Implementation of Programmatic Controls for 
Radiological Effluent Technical Specifications in the Administrative 
Controls Section of the Technical Specifications and the Relocation of 
Procedural Details of RETS to the Offsite Dose Calculation Manual or to 
the Process Control Program.''
    Date of issuance: September 8, 1999.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 121.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 19, 1999 (64 FR 
27324).
    The June 21, 1999, supplement provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination or expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 8, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: June 7, 1999, as supplemented by letters 
dated June 24 and August 24, 1999.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 2.0, ``Safety Limits and Limiting Safety System 
Settings,'' TS 3.2.5, ``DNB [Departure from Nucleate Boiling] 
Parameters,'' and the associated Bases, and Administrative Controls 
Section 6.9.1.6, ``Core Operating Limits Report [(COLR)],'' by 
relocating cycle-specific reactor coolant system-related parameter 
limits from the TSs to the COLR.
    Date of issuance: September 2, 1999.
    Effective date: September 2, 1999, to be implemented within 30 
days.
    Amendment Nos.: Unit 1--115; Unit 2--103.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 14, 1999 (64 FR 
38036).
    The August 24, 1999, supplement provided revised TS pages and 
clarifying information that was within the scope of the original 
Federal Register notice and did not change the staff's initial proposed 
no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated September 2, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.

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

    Date of application for amendments: September 4, 1998, as 
supplemented by letter dated November 25, 1998.
    Brief description of amendments: Revises the licensing basis to 
credit containment pressure in excess of atmospheric pressure in the 
analysis for Emergency Core Cooling Systems pump.
    Date of issuance: September 3, 1999.
    Effective date: As of date of issuance, to be incorporated into the 
Final Safety Analysis Report (FSAR) with the next update.
    Amendment Nos.: 261 and 220.
    Facility Operating License Nos. DPR-52 and DPR-68: Amendments 
approves changes to the FSAR.
    Date of initial notice in Federal Register: September 23, 1998 (63 
FR 5093). The November 25, 1998 supplemental letter did not change the 
original proposed no significant hazards determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 3, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Athens Public Library, 405 E. 
South Street, Athens, Alabama 35611.

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

    Date of application for amendments: January 15, 1999 (TS 98-09).

[[Page 51354]]

    Brief description of amendments: The amendments relocate seismic 
instrumentation requirements from the Technical Specifications to the 
Technical Requirements Manual.
    Date of issuance: September 7, 1999.
    Effective date: As of the date of issuance to be implemented no 
later than 45 days after issuance.
    Amendment Nos.: Unit 1--245; Unit 2--236.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6712).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 7, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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 24, 1999, as supplemented by letter 
dated July 9, 1999.
    Brief description of amendments: The amendments remove several 
cycle-specific parameter limits from the Technical Specifications 
(TSs). These parameter limits are added to the Core Operating Limits 
Report (COLR). Appropriate references to the COLR are inserted in the 
affected TSs. In addition, the core safety limit curves are replaced 
with safety limits more directly applicable to the fuel and fuel 
cladding fission product barriers.
    The affected TSs are: (1) TS 2.0, ``Safety Limits (Sls),'' (2) TS 
3.3.1, ``Reactor Trip System Instrumentation Setpoints,'' (3) TS 3.4.1, 
``RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling 
(DNB) Limits,'' and (4) TS 5.6.5, ``Core Operating Limits Report.''
    Date of issuance: August 30, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: 67 and 67.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35213) and July 28, 1999, (64 FR 40908).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 30, 1999.
    No significant hazards consideration comments received: No
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.

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

    Date of amendment request: May 14, 1999.
    Brief description of amendments: The amendments change the licenses 
to accurately reflect the new corporate name of the current licensee, 
``TXU Electric Company'' in Facility Operating Licenses NPF-87 and NPF-
89 for CPSES, Units 1 and 2, respectively.
    Date of issuance: August 31, 1999.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1--Amendment No. 68; Unit 2--Amendment No. 68.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
change the Operating Licenses.
    Date of initial notice in Federal Register: June 30, 1999 (64 FR 
35213).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated August 31, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.

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

    Date of application for amendment: May 26, 1999.
    Brief description of amendment: The amendment revises the 
suppression pool water temperature surveillance requirements to specify 
monitoring the temperature every 5 minutes when performing testing that 
adds heat to the suppression pool. In addition, the amendment revises 
the requirement to check the suppression chamber water level and 
temperature from ``once per shift'' to ``daily'' and specifies that it 
is the average temperature that is checked.
    Date of Issuance: August 30, 1999.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 174.
    Facility Operating License No. DPR-28.: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 28, 1999 (64 FR 
40909).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated August 30, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

Yankee Atomic Electric Co., Docket No. 50-29, Yankee Nuclear Power 
Station (YNPS) Franklin County, Massachusetts

    Date of application for amendment: March 17, 1999.
    Brief description of amendment: Revises the Possession Only License 
by deleting License Condition 2.C.(10) related to the Fitness-For-Duty 
program.
    Date of issuance: August 27, 1999.
    Effective date: August 27, 1999.
    Amendment No.: 152.
    Facility Operating License No. DPR-3. Amendment revises the 
license.
    Date of initial notice in Federal Register: June 2, 1999 (64 FR 
29717).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated August 27, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Greenfield Community College, 
1 College Drive, Greenfield, Massachusetts 01301.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was

[[Page 51355]]

not time for the Commission to publish, for public comment before 
issuance, its usual 30-day Notice of Consideration of Issuance of 
Amendment, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By October 22, 1999, 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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any

[[Page 51356]]

hearing held would take place while the amendment is in effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, 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 the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

    For the Nuclear Regulatory Commission.

    Dated at Rockville, Maryland, this 15th day of September, 1999.
Elinor G. Adensam,
Acting Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 99-24573 Filed 9-21-99; 8:45 am]
BILLING CODE 7590-01-P