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

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)

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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

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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

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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