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 May 22,

[[Page 33104]]

1998, through June 5, 1998. The last biweekly notice was published on

June 3, 1998 (63 FR 30261).

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed no Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

the Commission's regulations in 10 CFR 50.92, this means that operation

of the facility in accordance with the proposed amendment would not (1)

involve a significant increase in the probability or consequences of an

accident previously evaluated; or (2) create the possibility of a new

or different kind of accident from any accident previously evaluated;

or (3) involve a significant reduction in a margin of safety. The basis

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

determination. Any comments received within 30 days after the date of

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

expiration of the 30-day notice period. However, should circumstances

change during the notice period such that failure to act in a timely

way would result, for example, in derating or shutdown of the facility,

the Commission may issue the license amendment before the expiration of

the 30-day notice period, provided that its final determination is that

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

before action is taken. Should the Commission take this action, it will

publish in the Federal Register a notice of issuance and provide for

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

Written comments may be submitted by mail to the Chief, Rules and

Directives Branch, Division of Administration Services, Office of

Administration, U.S. Nuclear Regulatory Commission, Washington, DC

20555-0001, and should cite the publication date and page number of

this Federal Register notice. Written comments may also be delivered to

Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,

Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of

written comments received may be examined at the NRC Public Document

Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The

filing of requests for a hearing and petitions for leave to intervene

is discussed below.

By July 17, 1998, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

proceeding and who wishes to participate as a party in the proceeding

must file a written request for a hearing and a petition for leave to

intervene. Requests for a hearing and a petition for leave to intervene

shall be filed in accordance with the Commission's ``Rules of Practice

for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested

persons should consult a current copy of 10 CFR 2.714 which is

available at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC and at the local public

document room for the particular facility involved. If a request for a

hearing or petition for leave to intervene is filed by the above date,

the Commission or an Atomic Safety and Licensing Board, designated by

the Commission or by the Chairman of the Atomic Safety and Licensing

Board Panel, will rule on the request and/or petition; and the

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

As required by 10 CFR 2.714, a petition for leave to intervene

shall set forth with particularity the interest of the petitioner in

the proceeding, and how that interest may be affected by the results of

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

following factors: (1) the nature of the petitioner's right under the

Act to be made a party to the proceeding; (2) the nature and extent of

the petitioner's property, financial, or other interest in the

proceeding; and (3) the possible effect of any order which may be

entered in the proceeding on the petitioner's interest. The petition

should also identify the specific aspect(s) of the subject matter of

the proceeding as to which petitioner wishes to intervene. Any person

who has filed a petition for leave to intervene or who has been

admitted as a party may amend the petition without requesting leave of

the Board up to 15 days prior to the first prehearing conference

scheduled in the proceeding, but such an amended petition must satisfy

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

scheduled in the proceeding, a petitioner shall file a supplement to

the petition to intervene which must include a list of the contentions

which are sought to be litigated in the matter. Each contention must

consist of a specific statement of the issue of law or fact to be

raised or controverted. In addition, the petitioner shall provide a

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

contention and on which the petitioner intends to rely in proving the

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

aware and on which the petitioner intends to rely to establish those

facts or expert opinion. Petitioner must provide sufficient information

to show that a genuine dispute exists with the applicant on a material

issue of law or fact.

Contentions shall be limited to matters within the scope of the

amendment under consideration. The contention must be one which, if

proven, would entitle the petitioner to relief. A petitioner who fails

to file such a supplement which satisfies these requirements with

respect to at least one contention will not be permitted to participate

as a party.

Those permitted to intervene become parties to the proceeding,

subject to any limitations in the order granting leave to intervene,

and have the opportunity to participate fully in the conduct of the

hearing, including the opportunity to present evidence and cross-

examine witnesses.

If a hearing is requested, the Commission will make a final

determination on the issue of no significant hazards consideration. The

final determination will serve to decide when the hearing is held.

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

request for a hearing. Any hearing held would take place after issuance

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

A request for a hearing or a petition for leave to intervene must

be filed with the Secretary of the Commission, U.S. Nuclear Regulatory

Commission,

[[Page 33105]]

Washington, DC 20555-0001, Attention: Rulemakings and Adjudications

Staff, or may be delivered to the Commission's Public Document Room,

the Gelman Building, 2120 L Street, NW., Washington DC, by the above

date. A copy of the petition should also be sent to the Office of the

General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC

20555-0001, and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for a hearing will

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

petition and/or request should be granted based upon a balancing of

factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

For further details with respect to this action, see the

application for amendment which is available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document room for

the particular facility involved.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden

Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units

1 and 2, Rock Island County, Illinois

Date of application for amendment request: May 18, 1998.

Description of amendment request: Change various technical

specification (TS) values to conservatively reflect design values.

These TS values affect: (1) 125/250 volts direct current (Vdc)

electrolyte temperature; (2) control rod drive accumulator pressure;

(3) standby liquid control solution temperature; (4) ultimate heat sink

minimum water level; (5) shutdown suppression chamber level (Quad

Cities only); and (6) degraded voltage setpoint (Quad Cities only).

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Does the change involve a significant increase in the probability

or consequences of an accident previously evaluated?

The change does not involve a significant increase in the

probability or consequences of an accident previously evaluated. The

proposed changes to certain Technical Specification acceptance values

are conservative and serve to ensure operability of equipment important

to safety. By ensuring equipment availability, the probability or

consequences of an accident previously evaluated are not increased. In

addition, the proposed changes have no impact on any initial condition

assumptions for accident scenarios. Onsite or offsite dose consequences

resulting from an event previously evaluated are not affected by this

proposed amendment request.

Accordingly, there is no significant change in the probability or

consequences of an accident previously evaluated.

Does the change create the possibility of a new or different kind

of accident from any accident previously evaluated?

The proposed amendment does not create the possibility of a new or

different kind of accident from any accident previously evaluated. The

proposed license amendment provides changes in certain Technical

Specification values to restore margin and ensure equipment

operability. Each proposed change is conservative with respect to

current requirements. The proposed amendment does not involve any plant

physical changes that would create the possibility of a new or

different kind of accident from any accident previously evaluated.

Therefore, the proposed amendment does not create the possibility

of a new or different kind of accident from any accident previously

evaluated.

Does the change involve a significant reduction in a margin of

safety?

The proposed change does not involve a significant reduction in a

margin of safety. In fact, the proposed changes restore margin and

ensure equipment operability. Since the changes maintain the necessary

level of system reliability, they do not involve a significant

reduction in the margin of safety.

Therefore, the change does not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: for Dresden, Morris Area

Public Library District, 604 Liberty Street, Morris, Illinois 60450;

for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon,

Illinois 61021.

Attorney for licensee: Michael I. Miller, Esquire; Sidley and

Austin, One First National Plaza, Chicago, Illinois 60603.

NRC Project Director: Stuart A. Richards.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego County, New York

Date of application for amendment: May 2, 1998, as supplemented May

21, and 23 (three letters), 1998.

Brief description of amendment: This amendment changed Technical

Specification (TS) 3/4.6.2, ``Protective Instrumentation,'' and its

associated Bases to reflect modifications to the initiation

instrumentation for the Control Room Air Treatment System. It also

changed TS 3.2.4a, ``Reactor Coolant Activity,'' and added an

additional condition to the operating license.

Date of issuance: May 23, 1998.

Effective date: As of the date of issuance to be implemented prior

to resumption of power operation.

Amendment No.: 161.

Facility Operating License No. DPR-63: Amendment revises the

Technical Specifications. Public comments requested as to proposed no

significant hazards consideration: Yes (63 FR 27601 dated May 19, 1998.

The notice recognized the existence of exigent circumstances pursuant

to 10 CFR 50.91(a)(6) and provided an opportunity to submit comments on

the Commission's proposed no significant hazards consideration

determination. The notice published May 19, 1998, also provided for an

opportunity to request a hearing by June 1, 1998 (this will be

corrected to June 18, 1998, by a notice to be published in the near

future), but indicated that if the Commission makes a final no

significant hazards consideration determination, any such hearing would

take place after issuance of the amendment. Subsequent to publishing

the notice, and due to schedule improvements which have occurred at the

plant, the Commission has determined that the amendment should be

issued on an emergency basis pursuant to 10 CFR 50.91(a)(5). The

Commission's related evaluation of the amendment, finding of emergency

circumstances, consultation with the State of New York, and final no

significant hazards consideration determination are contained in a

Safety Evaluation date May 23, 1998.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State

[[Page 33106]]

University of New York, Oswego, New York 13126.

Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &

Strawn, 1400 L Street, NW, Washington, DC 20005-3502.

NRC Project Director: S. Singh Bajwa.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of amendment request: May 14, 1998.

Description of amendment request: The proposed amendment would

change the Technical Specifications (TSs) for the Reactor Protection

System (RPS) and the Engineered Safety Features Actuation System

(ESFAS) instrumentation by restricting the time most RPS and ESFAS

actuation channels can be in the bypass position to 48 hours. The

current TSs have no time limit. The proposed amendment would also

modify the TS action requirements and the channel calibration

requirements for the loss of turbine load reactor trip function, and

the channel calibration requirements for the wide range logarithmic

neutron flux monitors; add a note to exclude the neutron detectors from

the channel calibration requirements; correct a reference to a TS

surveillance requirement; and correct errors that have been identified.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to restrict the time most of the reactor

protection or engineered safety feature actuation channels can be in

the bypass position to 48 hours, from an indefinite period of time, has

no effect on the design of the Reactor Protection System (RPS) or the

Engineered Safety Feature Actuation System (ESFAS), and does not affect

how these systems operate. In addition, this will minimize the

susceptibility of these systems to the remote possibility of fault

propagation between channels. The pressurizer high pressure reactor

protection channels will not be required to be placed in the tripped

condition after 48 hours. A failed pressurizer high pressure channel

will be allowed to remain in the bypassed condition for up to 30 days.

If the failed pressurizer high pressure channel was placed in the

tripped condition, and then a high failure of another pressurizer high

pressure channel occurred, the reactor would trip and both pressurizer

power operated relief valves (PORVs) would open, resulting in an

undesired loss of primary coolant. Limiting the time that a failed

pressurizer high pressure reactor protection channel can be in bypass

to 30 days will minimize the risk of the inadvertent opening of both

PORVs, as well as the risk associated with fault propagation between

channels. These systems will still function as designed to mitigate

design basis accidents. Therefore, this change does not significantly

increase the probability or consequences of an accident previously

evaluated.

The proposed change to increase the time a second RPS or ESFAS

channel can be removed from service (from 2 hours to 48 hours),

provided one of the inoperable channels is placed in the tripped

condition, has no effect on the design of the RPS or ESFAS and does not

affect how these systems operate. These systems will still function as

designed to mitigate design basis accidents.

However, one of the proposed changes will allow two pressurizer

pressure reactor protection channels to be removed from service (one

channel in the tripped condition and one channel in the bypassed

condition) for 48 hours instead of the current 2 hour time limit. With

a pressurizer pressure channel in the tripped condition, the high

failure of a second pressurizer pressure channel would initiate a

reactor trip, open both pressurizer PORVs, and cause an undesired loss

of primary coolant. Thus, this change will increase the probability of

occurrence of a previously evaluated accident (FSAR [Final Safety

Analysis Report] Section 14.6.1--Inadvertent Opening of a Pressurized

Water Reactor Pressurizer Pressure Relief Valve). However, since this

configuration will only be allowed for an additional 46 hours, the

increase in the probability of occurrence of a previously evaluated

accident will be limited to an acceptable value. Therefore, this change

does not significantly increase the probability or consequences of an

accident previously evaluated.

The proposed change to apply a more restrictive action statement to

the loss of turbine load reactor trip function has no effect on the

design of this trip function and does not affect how this trip function

operates. Also, this trip function is not assumed to operate to

mitigate any design basis accident.

Therefore, this change does not significantly increase the

probability or consequences of accident previously evaluated.

The proposed change to require a channel calibration every 18

months for the loss of turbine load reactor trip function and for the

wide range logarithmic neutron flux monitors has no effect on the

design of either the loss of turbine load reactor trip function or the

wide range logarithmic neutron flux monitors. Also, neither of these

are assumed to operate to mitigate any design basis accident.

Therefore, this change does not significantly increase the probability

or consequences of an accident previously evaluated.

The proposed change to exclude the neutron detectors from the

channel calibration requirement has no effect on the design of the

neutron detectors and has no significant effect on how these detectors

operate. The detectors are passive devices with minimal drift. In

addition, slow changes in the sensitivity of the linear power range

flux detectors is compensated for by performing the daily calorimetric

calibration and the monthly calibration using the incore detectors.

These detectors will still function as designed to mitigate design

basis accidents. Therefore, this change does not significantly increase

the probability or consequences of an accident previously evaluated.

The proposed change to correct the surveillance requirement

referenced in an action statement has no effect on the design of the

ESFAS and does not affect how this system operates. The ESFAS will

still function as designed to mitigate design basis accidents.

Therefore, this change does not significantly increase the probability

or consequences of an accident previously evaluated.

The proposed change to add a reference to the reactor coolant pump

low speed reactor trip function to a note that states this trip may be

bypassed when [less than] 5 [percent] power, and that the bypass must

be automatically removed when [greater than or equal to] 5 [percent]

power will not effect this reactor trip function. This bypass

capability currently exists in the design of the Millstone Unit No. 2

RPS, and is the same bypass feature referenced for the reactor coolant

flow low reactor trip function. Both of these reactor trip functions

provide protection for a reduction in RCS [Reactor Coolant System]

flow. The addition of this note will not result in any technical change

to the Millstone Unit No. 2 RPS. The RPS will continue to function as

before. Therefore, this change does not significantly increase the

probability or consequences of an accident previously evaluated.

[[Page 33107]]

The proposed change to correct the power level high trip setpoint

on Technical Specification Page 2-4 will not result in any change to

the actual plant setpoint for this RPS trip function. As a result of

this proposed change, the setpoint listed on Page 2-4 will agree with

the setpoint previously approved by the NRC, and currently used by the

RPS. The change has no effect on the design of the RPS and does not

affect how this system operates. Therefore, this change does not

significantly increase the probability or consequences of an accident

previously evaluated.

The information added to the Bases of the Technical Specifications

to provide a discussion of how the RPS and ESFAS are affected by the

proposed changes, the effect the action statements have on the

operation of the RPS and ESFAS, and to discuss the impact of

surveillance testing on RPS operability will have no effect on

equipment operation. The RPS and ESFAS will continue to function as

designed to mitigate design basis accidents. Therefore, this change

does not significantly increase the probability or consequences of an

accident previously evaluated.

Thus, this License Amendment Request does not impact the

probability of an accident previously evaluated nor does it involve a

significant increase in the consequences of an accident previously

evaluated.

2. Create the possibility of a new or different kind of accident

from any accident previously evaluated.

The proposed changes do not alter the plant configuration (no new

or different type of equipment will be installed) or require any new or

unusual operator actions. They do not alter the way any structure,

system, or component functions and do not alter the manner in which the

plant is operated. The proposed changes do not introduce any new

failure modes. They will not alter assumptions made in the safety

analysis and licensing basis. The RPS and the ESFAS will still function

as designed to mitigate design basis accidents.

Therefore, these changes do not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

The proposed changes will not reduce the margin of safety since

they have no impact on any safety analysis assumption. The proposed

changes do not decrease the scope of equipment currently required to be

operable or subject to surveillance testing, nor do the proposed

changes affect any instrument setpoints or equipment safety functions.

The effectiveness of Technical Specifications will be maintained

since the changes will not alter the operation of any RPS or ESFAS

function. In addition, most of the changes are consistent with the

Calvert Cliffs RPS and ESFAS Technical Specifications mode provided in

Enclosure 3 of the NRC correspondence dated April 16, 1981 (R. A. Clark

letter to W. G. Counsil, Evaluation of the Reactor Protection System

Inoperable Channel Condition at Millstone Nuclear Power Station, Unit

No. 2, dated April 16, 1981) and the new, improved Standard Technical

Specifications (STS) for Combustion Engineering plants (NUREG-1432).

Therefore, there is no significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut

Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear

Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,

Connecticut.

NRC Deputy Director: Phillip F. McKee.

Power Authority of The State of New York, Docket No. 50-286, Indian

Point Nuclear Generating Unit No. 3, Westchester County, New York

Date of amendment request: June 25, 1997.

Description of amendment request: The proposed amendment would

change the Indian Point 3 Technical Specifications to allow the use of

zirconium alloy or stainless steel filler rods in fuel assemblies to

replace failed or damaged fuel rods.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Consistent with the criteria of 10 CFR 50.92, the enclosed

application is judged to involve no significant hazards based on the

following information:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident previously

analyzed?

Response: The proposed changes modify the technical specification

only to the extent that the reconstitution is recognized as acceptable

under limited circumstances. Reconstitution is limited to substitution

of zirconium alloy or stainless steel filler rods, and must be in

accordance with approved applications of fuel rod configurations.

Although these changes permit reconstitution to occur without the need

for a specific technical specification change, use of an approved

methodology is required prior to its application. Since the changes

will allow substitution of filler rods for leaking, potentially leaking

rods or damaged rods, the changes may actually reduce the radiological

consequences of an accident. It is noted that the specific changes

requested in this letter have previously been found acceptable by the

NRC in GL [Generic Letter] 90-02, Supplement 1. For these reasons, we

conclude that the changes will not involve a significant increase in

the probability or consequences of an accident previously evaluated.

(2) Does the proposed license amendment create the possibility of a

new or different kind of accident from any previously evaluated?

Response: The proposed changes will not create the possibility of a

new or different kind of accident from any accident previously

evaluated because they will only affect the assembly configuration and

can only be implemented if demonstrated to meet current plant

requirements in accordance with an NRC-approved methodology. The other

aspects of plant design, operation limitations, and responses to events

will remain unchanged. It is noted that the changes have previously

been determined acceptable by the NRC in GL 90-02, Supplement 1.

(3) Does the proposed amendment involve a significant reduction in

a margin of safety?

Response: The proposed change will not involve a reduction in a

margin of safety because the changes can only be implemented if

demonstrated to meet current plant requirements in accordance with an

NRC-approved methodology. It is noted that the changes have previously

been determined acceptable by the NRC in GL 90-02, Supplement 1.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 50.92(c) are

satisfied.

[[Page 33108]]

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10601.

Attorney for licensee: Mr. David Blabey, 10 Columbus Circle, New

York, New York 10019.

NRC Project Director: S. Singh Bajwa.

Public Service Electric & Gas Company, Docket No. 50-354, Hope Creek

Generating Station, Salem County, New Jersey

Date of amendment request: April 28, 1998.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) 3.4.2.1 to replace the plus or

minus 1 percent setpoint tolerance limit for safety/relief valves

(SRVs) with a plus or minus 3 percent setpoint tolerance limit. In

addition, the proposed amendment would revise TS 4.4.2.2 to state that

all SRVs must be certified to be within plus or minus 1 percent of the

TS setpoint prior to returning the valves to service.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed TS revisions involve: (1) no significant hardware

changes; (2) no significant changes to the operation of any systems or

components in normal or accident operating conditions; and (3) no

changes to existing structures, systems, or components. Therefore these

changes will not increase the probability of an accident previously

evaluated.

These proposed changes were developed in accordance with the

provisions contained in an NRC Safety Evaluation Report, dated 3/8/93,

for the ``BWR Owners Group Inservice Pressure Relief Technical

Specification [Revision] Licensing Topical Report'', NEDC-31753P as

described in General Electric report NEDC-32511P, ``Safety Review for

Hope Creek [Generating Station] Safety/Relief Valve Tolerance

Analyses''. Since the plant systems associated with these proposed

changes will still be capable of: (1) meeting all applicable design

basis requirements; and (2) retain the capability to mitigate the

consequences of accidents described in the HC [Hope Creek] UFSAR

[Updated Final Safety Analysis Report], the proposed changes were

determined to be justified. Therefore, these changes will not involve a

significant increase in the consequences of an accident previously

evaluated.

(2) The proposed change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Establishment of the [plus or minus] 3 [percent] SRV setpoint

tolerance limit will not adversely impact the operation of any safety

related component or equipment. Since the proposed changes involve: (1)

no significant hardware changes; (2) no significant changes to the

operation of any systems or components; and (3) no changes to existing

structures, systems, or components, there can be no impact on the

occurrence of any accident. These proposed changes were developed in

accordance with the provisions contained in an NRC Safety Evaluation

Report, dated 3/8/93, for the ``BWR Owners Group Inservice Pressure

Relief Technical Specification [Revision] Licensing Topical Report'',

NEDC-31753P as described in General Electric report NEDC-32511P,

``[Safety Review for Hope Creek Generating Station] Safety/Relief Valve

Tolerance Analyses''. Furthermore, there is no change in plant testing

proposed in this change request which could initiate an event.

Therefore, these changes will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

(3) The proposed change does not involve a significant reduction in

a margin of safety.

Establishment of the [plus or minus] 3 [percent] SRV setpoint

tolerance limit will not adversely impact the operation of any safety

related component or equipment. General Electric analyses performed for

Hope Creek and contained in General Electric report NEDC-32511P,

``[Safety Review for Hope Creek Generating Station] Safety/Relief Valve

Tolerance Analyses,'' concluded that there is no significant impact on

fuel thermal limits, no significant impact on safety related systems,

structures or components, and no significant impact on the accident

analyses associated with the proposed changes. Therefore, the changes

contained in this request do not result in a significant reduction in a

margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.

NRC Project Director: Robert A. Capra.

Southern Nuclear Operating Company, Inc., Georgia Power Company,

Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,

City of Dalton, Georgia, Docket Nos. 50-424 and 50-425, Vogtle Electric

Generating Plant, Units 1 and 2, Burke County, Georgia

Date of amendment request: May 8, 1998.

Description of amendment request: The proposed amendments would

change the Vogtle Electric Generating Plant (VEGP) Technical

Specification (TS) 5.5.7, ``Reactor Coolant Pump Flywheel Inspection

Program,'' to provide an exception to the examination requirements of

Regulatory Position C.4.b of Regulatory Guide (RG) 1.14, Revision 1,

August 1975.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the probability

or consequences of an accident previously evaluated.

The safety function of the RCP [reactor coolant pump] flywheel is

to provide sufficient rotational inertia to ensure reactor coolant flow

through the core during coastdown following a loss of offsite power and

subsequent reactor trip. FSAR [Final Safety Analysis Report] Chapter 15

analysis for a complete loss of forced reactor coolant flow

demonstrates that the reactor trip together with the flow sustained by

the inertia of the RCP impeller will be sufficient to prevent the most

limiting fuel assembly from exceeding the DNBR [departure from nucleate

boiling ratio] limits.

The maximum mechanical loading on the RCP motor flywheel results

from overspeed following a LOCA [loss-of-coolant accident]. The

analysis presented in WCAP-14535A demonstrates that the revised

inspection program proposed by this license amendment will ensure the

integrity of the RCP flywheels will be maintained.

[[Page 33109]]

Based upon the findings of WCAP-14535A, the ability of the RCP

flywheel to perform its intended safety function will be unaffected by

the license amendment and the FSAR Chapter 15 analysis will remain

valid. Therefore, these proposed changes do not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

(2) Operation of the facility in accordance with the proposed

amendment would not create the possibility of a new or different kind

of accident from any accident previously evaluated.

The proposed license amendment will not change the physical plant

configuration nor the modes of operation of any plant equipment. Based

upon the results of WCAP-14535A, no new failure mechanism will be

introduced by the revised RCP flywheel inspection program. Therefore,

the proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

(3) Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The operating limits and functional capabilities of the affected

systems, structures, and components will be unchanged by the proposed

amendment. The results of the RCP flywheel inspections performed

throughout the industry and at VEGP have identified no indications

which would affect its integrity. As presented in WCAP-14535A, detailed

stress analysis and risk assessments have been completed with the

results indicating that there would be no change in the probability of

failure for RCP flywheels if all inspections were eliminated.

Therefore, these changes do not involve a significant reduction in a

margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Burke County Public Library,

412 Fourth Street, Waynesboro, Georgia.

Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,

NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,

Georgia.

NRC Project Director: Herbert N. Berkow.

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,

Unit 1, Rhea County, Tennessee

Date of amendment request: May 6, 1998.

Description of amendment request: The proposed amendment would

replace the two percent penalty addressed in surveillance requirement

(SR) 3.2.1.2(a) with a burnup-dependent factor to be specified in the

Watts Bar Core Operating Limits Report (COLR). Specifically, the

following changes are being proposed:

1. SR 3.2.1.2(a) and its associated BASES will have the phrase ``by

a factor of 1.02'' deleted and replaced with the phrase ``by the

appropriate factor specified in the COLR.''

2. Technical Specification (TS) Section 5.9.5(b)(3) would be

updated to reference the revised WCAP (10216-P-A, Revision 1A, 1994)

that details the analytical methods utilized for the new penalty

factor.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

A. The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously evaluated.

The proposed change involves only the manner in which the penalty

factors for FQ(Z) would be specified (i.e., burnup-dependent

factor specified in the Core Operating Limits Report [COLR] versus a

constant factor specified in the TS). This is simply used to account

for the fact that FQ C(Z) may increase between

surveillance intervals. These penalty factors are not assumed in any of

the initiating events for the accident analyses. Therefore the proposed

change will have no effect on the probability of any accidents

previously evaluated. The penalty factors specified in the COLR will be

calculated using NRC-approved methodology and will continue to provide

an equivalent level of protection as the existing TS requirement.

Therefore, the proposed change will not affect the consequences of any

accident previously evaluated.

B. The proposed amendment does not create the possibility of a new

or different kind of accident from any accident previously evaluated.

The proposed change does not involve a physical alteration to the

plant (no new or different kind of equipment will be installed) or

alter the manner in which the plant would be operated. Thus, this

change does not create the possibility of a new or different kind of

accident from any previously evaluated.

C. The proposed amendment does not involve a significant reduction

in a margin of safety.

The proposed change will continue to ensure that potential

increases in FQ C(Z) over a surveillance interval

will be properly accounted for. The penalty factors will be calculated

using an NRC-approved methodology. Therefore, the proposed change will

not involve a reduction in margin of safety.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902.

NRC Project Director: Frederick J. Hebdon.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: May 1, 1998.

Description of amendment request: The proposed amendment would make

several editorial changes to the Administrative Controls section of the

Technical Specifications. The changes include revisions due to

organizational changes, quality assurance changes, editorial changes,

and typographical corrections.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Will the proposed changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

The administrative change proposed herein will have no effect on

plant hardware, plant design, safety limit setting or plant system

operation and therefore do[es] not modify or add any initiating

parameters that would significantly increase the probability or

consequences of any previously analyzed accident. The proposed

amendment changes the reference to the VYNPS QA program and makes other

[[Page 33110]]

administrative changes, such as title changes and correction/

clarification of errors. Therefore, there is no increase in the

probability or consequence of an accident previously evaluated.

2. Will the proposed changes create the possibility of a new or

different kind of accident from any accident previously evaluated?

This change does not affect any equipment nor does it involve any

potential initiating events that would create any new or different kind

of accident. The proposed change involves [ ] wording changes in the

Technical Specifications identifying the name of the QA program and

makes other administrative changes, such as title changes and

corrective/clarification of errors. Therefore no new or different kind

of accident has been introduced.

3. Will the proposed changes involve a significant reduction in a

margin of safety?

This change does not affect any equipment involved in potential

initiating events or safety limits. The proposed change has no

significant impact on margin of safety, as it is comprised of only

administrative changes.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Project Director: Cecil O.Thomas.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,

North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of amendment request: September 1, 1995, as supplemented April

8, 1996, April 22, 1996, April 23, 1996, November 18, 1997, February 9,

1998, March 25, 1998 and May 5, 1998. This notice supersedes the

Federal Register notice of September 27, 1995 (60 FR 49949)

Description of amendment request: The originally (September 1,

1995) proposed changes to the Technical Specifications (TS) would

permit a single outage of up to 14 days for each emergency diesel

generator (EDG) once every 18 months in order to perform preventive

maintenance. The amended request will permit a single outage of up to

14 days for each EDG for any reason; TS change to incorporate a

Configuration Risk Management Program (CRMP) in the Administrative

Section in the TS, in support of the previous submittal for the 14-day

Allowed Outage Time (AOT) for the EDGs and would permit an increase in

the TS maintenance interval of the EDG from 18 to 24 months, based on

the recommendation from the EDG owners group (Fairbanks Morse Owners

Group).

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration. Specifically, operation of North Anna Power Station in

accordance with the proposed Technical Specification changes will not:

a. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

A probabilistic safety analysis (PSA) has been performed which

demonstrates that a 14-day AOT for each EDG, results in a small change

in core damage frequency assuming adequate compensatory measures are in

place. The compensatory measures include requirements that the other

EDGs, off-site power supply, and the alternate A.C. diesel (AAC DG) be

operable whenever the action statement is entered.

The effect of the proposed change has been calculated to be an

increase in core damage frequency of approximately 1 E-6 per year from

the baseline core damage frequency of 4.1 E-5. Considering that credit

was not taken for the AAC DG previously in the IPE nor was the AAC DG

specified in Technical Specifications, the proposed changes remain

bounded by the core damage frequency identified in the Individual Plant

Examination.

Credit for the AAC DG was previously not taken nor was the AAC DG

previously included in the Technical Specifications. Furthermore, the

probabilistic safety analysis (PSA) demonstrates that the increase in

core damage frequency due to extending the EDG AOT of a 14-day period

is not significant as long as the AAC DG is operable to act as a source

of emergency power to replace the EDG. The period of time during which

the EDG is unavailable is short enough to limit the impact of using the

manually operated AAC DG as a replacement for the automatically

operated EDG.

The plant design and operation are not changed by the incorporation

of a CRMP into the Administrative Section of Technical Specifications.

Further, with the proposed change to the preventive maintenance

interval, the EDG reliability remains adequate to perform its function

of supporting accident mitigation equipment with emergency electrical

power.

Therefore, neither the probability of occurrence nor the

consequences of an accident or malfunction of equipment important to

safety previously evaluated in the safety analysis report are increased

due [to] the proposed changes to permit a 14-day allowed outage time

and a 24 month preventive maintenance interval for the EDGs.

b. Create the possibility of a new or different kind of accident

from any accident previously evaluated.

No new initiators are defined as a result of a review of the PSA

model. The proposed Technical Specifications changes only modify the

AOT of an EDG. The UFSAR [Updated Final Safety Analysis Report]

accidents are analyzed assuming that the EDG is the worst single

failure. This assumption is more severe than the proposed Technical

Specifications changes, which [replace] the EDG with the AAC DG.

Similarly, the PSA performed to evaluate the proposed Technical

Specifications changes considered all of the initiating events defined

for the PSA performed for the Individual Plant Examination. No new

initiators were defined as a result of a review of the PSA model.

Adding the CRMP and changing the EDG preventive maintenance

interval in the Technical Specifications does not change any method of

operation or create any new modes of operation or accident precursors.

Therefore, it is concluded that no new or different kind of

accident or malfunction from any previously evaluated has been or will

be created by the proposed changes to permit a 14-day allowed outage

time and a 24 month preventive maintenance interval for the EDGs.

c. The proposed Technical Specifications changes do not result in a

reduction in margin of safety as defined in the basis for any Technical

Specifications.

The PSA was performed to evaluate the concept of a one-time outage.

The results of the analyses show a small change in the core damage

frequency. As described above the proposed Technical Specifications

changes only modify the AOT of an EDG. Thus, operation with slightly

increased EDG unavailability due to maintenance is acceptable given the

operability of the AAC DG and the other EDG.

Incorporating the CRMP and changing the EDG preventive maintenance

interval in the Technical Specifications

[[Page 33111]]

does not affect any accident analysis assumptions or change any

Technical Specifications criteria.

Therefore, the margin of safety is not changed.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Attorney for licensee: Michael W. Maupin, Esq., Hunton and

Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,

Virginia 23219.

NRC Project Director: Gordon E. Edison, Acting.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: May 7, 1998.

Description of amendment request: Technical Specification 5.4,

``Fuel Storage,'' would be changed to increase the allowable mass of

uranium-235, per axial centimeter, for fuel storage in new fuel and

spent fuel storage racks. This change will allow use of new Siemens

heavy fuel assemblies.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change was reviewed in accordance with the provisions

of 10 CFR 50.92 to show no significant hazards exist. The proposed

change will not:

(1) Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The mass of the fuel assembly is increased by a small amount (30

pounds, or 2.4%), from that of the fuel assemblies now in the core.

Even with this increase, the load on the fuel handling equipment is

still well within design limits. Therefore, the probabilities of a fuel

handling accident inside containment (FHAIC) and the fuel handling

accident outside containment (FHAOC) are not changed.

The total core mass, with Siemens heavy fuel, is less than that

assumed in the original plant safety analysis. The proposed change does

not alter the plant configuration, operating set points, or overall

plant performance. The probability of other accidents is therefore not

changed.

Attachment 4 (of the application) shows that the consequences of a

fuel handling accident or a large break loss of coolant accident are

not significantly affected.

Any changes in the nuclear properties of the reactor core that may

result from a higher mass of fuel U235 per axial centimeter

will be analyzed and shown to meet acceptance criteria in the

appropriate reload analysis, which would be completed prior to use.

(2) Create the possibility of a new or different kind of accident

from any previously evaluated.

As discussed above, the only safety issue significantly affected by

the proposed change is the criticality analysis of the spent fuel

storage racks and new fuel storage racks. Since it has been

demonstrated that keff remains below the keff

acceptance criteria, no new or different accident would be created

through the use of fuel with up to 56.067 grams of U235 per

axial centimeter at the Kewaunee Nuclear Power Plant.

The proposed change does not alter the plant configuration,

operating set points, or overall plant performance and therefore does

not create a new or different kind of accident from any accident

previously evaluated.

(3) Involve a significant reduction in the margin of safety.

The criticality analysis in Reference 3 (of the application)

demonstrates that adequate margins to criticality can be maintained

with up to 56.067 grams of U235 per axial centimeter stored

in either the new fuel storage racks or the spent fuel storage racks.

The bounding cases of the analysis demonstrate that keff

remains less than 0.95 in the spent fuel storage racks and the new fuel

storage racks if flooded with unborated water. The bounding cases of

the analysis also demonstrate that keff remains less than

0.98 in the new fuel storage racks if moderated by optimally misted

moderator. Therefore, the 56.067 grams of U235 per axial

centimeter limit is acceptable for storage in both the new fuel storage

racks and the spent fuel storage racks.

Any changes in the nuclear properties of the reactor core that may

result from a higher mass of fuel U235 per axial centimeter

will be analyzed in the appropriate reload analysis to ensure

compliance with applicable reload considerations and requirements.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,

P.O. Box 1497, Madison, WI 53701-1497.

NRC Acting Project Director: Ronald R. Bellamy.

Wisconsin Electric Power Company, Docket No. 50-301, Point Beach

Nuclear Plant, Unit 2, Town of Two Creeks, Manitowoc County, Wisconsin

Date of amendment request: May 15, 1998 (NPL-98-0303).

Description of amendment request: The proposed amendment revises

the schedule for implementing the boron concentration changes related

to the planned conversion of Unit 2 to 18-month fuel cycles.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensee has

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendment will not result in a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes are administrative only. There are no physical

changes to the facility or its operation. All Limiting Conditions of

Operation, Limiting Safety System Settings, and Safety Limits specified

in the Technical Specification remain unchanged. Additionally, there

are no changes in the Quality Assurance Program, Emergency Plan,

Security Plan, and Operator Training and Requalification Program.

Therefore, an increase in the probability or consequences of an

accident previously evaluated cannot occur.

2. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendment will not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The proposed changes are administrative only. No changes to the

facility structures, systems and components or their operation will

result. The design and design basis of the facility remain unchanged.

The plant safety analyses remain current and accurate. No new or

different failure mechanisms are introduced. Therefore,

[[Page 33112]]

the possibility of a new or different kind of accident from any

accident previously evaluated is not introduced.

3. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendment does not involve a significant reduction in a

margin of safety.

The proposed [amendment is] administrative only. All safety margins

established through the design and facility license including the

Technical Specifications remain unchanged. Therefore, all margins of

safety are maintained.

The NRC staff has reviewed the licensee's analysis and, based on

this review, it appears that the three standards of 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,

and Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Project Director: Cynthia A. Carpenter.

Previously Published Notices of Consideration of Issuance of Amendments

to Facility Operating Licenses, Proposed no Significant Hazards

Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

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

Michigan

Date of amendment request: May 20, 1998 (NRC-98-0099).

Description of amendment request: The proposed amendment would

modify the scram discharge volume (SDV) vent and drain valve action

requirements to be consistent with those contained in NUREG-1433,

Revision 1, ``Standard Technical Specifications General Electric

Plants, BWR/4.''

Detroit Edison is requesting that this license amendment request be

processed in an exigent manner in accordance with 10 CFR 50.91(a)(6)

because delay in granting this amendment could lead to a plant

shutdown.

Date of publication of individual notice in Federal Register: May

28, 1998 (63 FR 29254).

Expiration date of individual notice: Comments: June 11, 1998;

hearing: June 29, 1998.

Local Public Document Room location: Monroe County Library System,

Ellis Reference and Information Center, 3700 South Custer Road, Monroe,

Michigan 48161.

Duke Energy Corporation, Docket Nos. 50-413 and 50-414, Catawba Nuclear

Station, Units 1 and 2, York County, South Carolina

Date of amendment request: May 22, 1998.

Description of amendment request: The proposed amendments would

revise Surveillance Requirement Section 4.4.3.3 of the Technical

Specifications. Section 4.4.3.3 currently requires that the emergency

power supply for the pressurizer heaters be demonstrated OPERABLE at

least once per 18 months by manually transferring power from the normal

to the emergency power supply. The licensee proposed to delete the

``manual'' requirement because the power supply transfer at the unit

was designed to be automatic. The proposed requirement is to verify

that required pressurizer heaters are capable of being powered from an

emergency power supply once per 18 months.

Date of publication of individual notice in Federal Register: June

1, 1998 (63 FR 29759).

Expiration date of individual notice: July 1, 1998.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego County, New York

Date of application for amendment: May 2, 1998.

Brief description of amendment: The amendment changes the Technical

Specifications 3/4.6.2, ``Protective Instrumentation,'' to reflect

modifications to the initiation instrumentation for the Control Room

Air Treatment system.

Date of publication of individual notice in Federal Register: May

19, 1998 (63 FR 27601).

Expiration date of individual notice: June 18, 1998.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego County, New York

Date of application for amendment: May 15, 1998 (two letters).

Brief description of amendment: The amendment changes

administrative sections of the Technical Specifications to reflect a

restructuring of upper management organization.

Date of publication of individual notice in Federal Register: June

2, 1998 (63 FR 30026).

Expiration date of individual notice: July 2, 1998.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: May 12, 1998.

Brief description of amendment request: These amendments relocate

certain requirements related to fire protection from the TSs to the

Updated Final Safety Analysis Report. The TS sections to be relocated

are: 3/4.3.7.9, Fire Detection Instrumentation; 3/4.7.6, Fire

Suppression Systems; 3/4.7.7, Fire Rated Assemblies; and 6.2.2e, Fire

Brigade Staffing. The amendments also replace License Condition 2.C.(6)

for Unit 1 and License Condition 2.C.(3) for Unit 2. These amendments

are consistent with the guidance of NRC Generic Letter (GL) 86-10,

``Implementation of Fire Protection Requirements,'' and GL 88-12,

``Removal of Fire Protection Requirements from Technical

Specifications.''

Date of publication of individual notice in Federal Register: May

21, 1998 (63 FR 28010).

Expiration date of individual notice: June 22, 1998.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

[[Page 33113]]

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna

Nuclear Power Plant, Wayne County, New York

Date of amendment request: March 31, 1997, as supplemented June 18,

1997, October 10, 1997, October 20, 1997, November 11, 1997, December

22, 1997, January 15, 1998, January 27, 1998, March 30, 1998, April 23,

1998, and April 27, 1998.

Brief description of amendment request: The proposed amendment

would revise the Ginna Station Improved Technical Specifications to

reflect a planned modification to the spent fuel pool storage racks.

Date of publication of individual notice in Federal Register: May

12, 1998 (63 FR 26213). This notice supersedes the March 31, 1997,

application published on April 30, 1997 (62 FR 23502).

Expiration date of individual notice: June 11, 1998.

Local Public Document Room Location: Rochester Public Library, 115

South Avenue, Rochester, New York 14610.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

10 CFR 51.12(b) and has made a determination based on that assessment,

it is so indicated.

For further details with respect to the action see (1) the

applications for amendment, (2) the amendment, and (3) the Commission's

related letter, Safety Evaluation and/or Environmental Assessment as

indicated. All of these items are available for public inspection at

the Commission's Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC, and at the local public document rooms for

the particular facilities involved.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,

Maryland

Date of application for amendments: January 31, 1997, as

supplemented February 13, February 28, March 25, April 16, August 19,

and September 29, 1997, January 22, March 17, April 8, April 21, 1998,

and May 22, 1998.

Brief description of amendments: The amendments revise the TS for a

reduction of the total reactor coolant system flow limit from 370,000

gallons per minute (gpm) to 340,000 gpm in support of increased steam

generator tube plugging.

Date of issuance: May 23, 1998.

Effective date: As of the date of issuance Unit 1 to be implemented

within 60 days and Unit 2 prior to startup from the spring 1999

refueling outage.

Amendment Nos.: 228 and 202.

Facility Operating License Nos. DPR-53 and DPR-69: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: February 26, 1997 (62

FR 8780).

The February 13, February 28, March 25, April 16, August 16, and

September 29, 1997, January 22, March 17, April 8, and April 21, 1998,

and May 22, 1998, letters provided clarifying information that did not

change the initial proposed no significant hazards consideration.

The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated May 23, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of application for amendments: July 18, 1997.

Brief description of amendments: The amendments revise the listed

design suppression chamber temperature of 200 deg.F to 220 deg.F and

the listed total water and steam volume of the reactor coolant system

from 18,670 cubic feet to 18,320 cubic feet, respectively.

Date of issuance: May 27, 1998.

Effective date: May 27, 1998.

Amendment Nos.: 195 and 225.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments

revise the facility's Technical Specifications.

Date of initial notice in Federal Register: August 27, 1997 (62 FR

45454).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 27, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of application for amendments: June 12, 1997, as supplemented

February 2, 1998. The February 2, 1998, submittal contained clarifying

information only and did not change the initial proposed no significant

hazards consideration or expand the scope of the original Federal

Register Notice.

Brief Description of amendments: The amendments consist of changes

to the Technical Specifications (TS) to revise the Limiting Condition

for Operation of the TS to limit the drywell average air temperature

rather than primary containment air temperature. Additionally, the

amendments require that the drywell average air temperature be

maintained less than or equal to 150 deg.F during plant operation. The

current primary containment average temperature limit is 135 deg.F.

Date of issuance: May 28, 1998.

Effective date: May 28, 1998.

Amendment Nos.: 196 and 226.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments

change the Technical Specifications.

Date of initial notice in Federal Register: August 27, 1997 (62 FR

45454) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated May 28, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

[[Page 33114]]

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of application for amendments: October 28, 1997

Brief Description of amendments: The amendments revise certain

instrumentation allowable values in the current technical

specifications to the Improved Technical Specifications format.

Date of issuance: May 28, 1998.

Effective date: May 28, 1998.

Amendment Nos.: 197 and 227. Facility Operating License Nos. DPR-71

and DPR-62: Amendments change the Technical Specifications.

Date of initial notice in Federal Register: December 31, 1997 (62

FR 68304)

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 28, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Carolina Power & Light Company, et al., Docket Nos. 50-325 & 50-324,

Brunswick Steam Electric Plant, Units 1 & 2, Brunswick County, North

Carolina

Date of amendment request: November 15, 1995.

Brief description of amendment: The amendments modify the channel

functional test interval in the Technical Specifications Surveillance

Requirements for the Electrical Protective Assemblies in the Reactor

Protection System.

Date of issuance: May 29, 1998.

Effective date: May 29, 1998.

Amendment No.: 198 and 228.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments

revise the Technical Specifications.

Date of initial notice in Federal Register: July 3, 1996 (61 FR

34887).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 29, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Carolina Power & Light Company, et al., Docket Nos. 50-325 & 50-324,

Brunswick Steam Electric Plant, Units 1 & 2, Brunswick County, North

Carolina

Date of amendment request: November 16, 1994, as supplemented by

letters dated February 14, 1995, and April 9, 1998.

Brief description of amendment: The amendments change the Technical

Specifications (TS) for Units 1 and 2 to revise the basis for removing

the suppression chamber water temperature monitoring instrumentation

requirements from the TS. This change is being processed in parallel

with the Improved Technical Specification conversion.

Date of issuance: May 29, 1998.

Effective date: May 29, 1998.

Amendment Nos.: 199 and 229.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments

revise the Technical Specifications.

Date of initial notice in Federal Register: January 4, 1995 (60 FR

497)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 29, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of application for amendments: April 4, 1996, as supplemented

January 24, 1997, March 31, 1997, April 2, 1997, April 14, 1997, March

24, 1998, and May 20, 1998.

Brief Description of amendments: The amendments modify Technical

Specifications (TS) 3.0.4, 4.0.3, and 4.0.4, and their associated Bases

in accordance with the guidance provided in Generic Letter 87-09,

``Sections 3.0 and 4.0 of the Standard Technical Specifications (STS)

on the Applicability of Limiting Conditions for Operation and

Surveillance Requirements.''

Date of issuance: June 2, 1998.

Effective date: June 2, 1998.

Amendment Nos.: 200 and 230.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments

change the Technical Specifications.

Date of initial notice in Federal Register: July 17, 1996 (61 FR

37297).

The supplemental submittals contained clarifying information only,

and did not change the initial no significant hazards consideration

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 2, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297

Carolina Power & Light Company, et al., Docket Nos. 50-325 & 50-324,

Brunswick Steam Electric Plant, Units 1 & 2, Brunswick County, North

Carolina

Date of amendment request: April 30, 1997, as supplemented October

28, 1997, and May 15, 1998.

Brief description of amendment: The amendments revise surveillance

requirements 4.7.2.b.2 and 4.7.2.c to require testing of the control

room emergency ventiliation system charcoal adsorber in accordance with

the American Society for Testing and Material D3803-1989, ``Standard

Test Method for Nuclear-Grade Activated Carbon.''

Date of issuance: June 2, 1998.

Effective date: June 2, 1998.

Amendment Nos.: 201 and 231.

Facility Operating License Nos. DPR-71 and DPR-62: Amendments

revise the Technical Specifications.

Date of initial notice in Federal Register: July 30, 1997 (62 FR

40846).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 2, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,

Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North

Carolina

Date of application for amendments: April 3, 1998

Brief description of amendments: The amendments revise the

specified total volume of the condensate storage tank capacity

requirements from 150,000 gallons to 228,200 gallons to ensure the Core

Spray System requirement of 50,000 gallons.

Date of issuance: June 5, 1998.

Effective date: June 5, 1998.

Amendment Nos.: 202 and 232.

[[Page 33115]]

Facility Operating License Nos. DPR-71 and DPR-62: Amendments

revise the facility's Technical Specifications.

Date of initial notice in Federal Register: May 6, 1998 (63 FR

25103).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 5, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

at Wilmington, William Madison Randall Library, 601 S. College Road,

Wilmington, North Carolina 28403-3297.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North

Carolina

Date of application for amendment: October 29, 1997.

Brief description of amendment: This amendment changes Technical

Specifications (TS) 3.8.1.1.a.3, 3.8.1.1.b.4, and 3.8.1.1.d.2 by

eliminating the plant shutdown requirements in these TS, and allowing

the applicable redundant feature TS to direct the plant shutdown when

required.

Date of issuance: May 22, 1998.

Effective date: May 22, 1998.

Amendment No.: 78.

Facility Operating License No. NPF-63: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: December 31, 1997 (62

FR 68305).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 22, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon

Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North

Carolina

Date of application for amendment: April 24, 1998, as supplemented

by letter dated May 15, 1998.

Brief description of amendment: This amendment revises TS 3.3.2,

``Engineered Safety Features Actuation System Instrumentation,'' such

that surveillance of the undervoltage relays may be performed without

entry into TS 3.0.3. Specifically, the change modifies Table 3.3-3 to

allow operation with more than one channel of the emergency bus

undervoltage relays inoperable.

Date of issuance: June 3, 1998.

Effective date: June 3, 1998.

Amendment No.: 79.

Facility Operating License No. NPF-63: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: May 4, 1998 (63 FR

24574).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 3, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Cameron Village Regional

Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois Docket Nos. STN

50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will

County, Illinois

Date of application for amendments: September 24, 1997.

Brief description of amendments: The amendments revise the

surveillance frequency for the turbine throttle valves and the turbine

governor valves from monthly to quarterly.

Date of issuance: May 26, 1998.

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 103 and 93.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revised the Technical Specifications.

Date of initial notice in Federal Register: March 11, 1998 (63 FR

11917).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 26, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: For Byron, the Byron Public

Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010;

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

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

Michigan

Date of application for amendment: January 28, 1998 (NRC-98-0006),

as supplemented on March 10, 1998 (NRC-98-0036).

Brief description of amendment: The amendment revises technical

specification surveillance requirement 4.4.3.2.2.a for the leak rate

test of the pressure isolation valves, extending it from the current

18-month interval to a 24-month interval.

Date of issuance: May 28, 1998.

Effective date: May 28, 1998, with full implementation within 90

days.

Amendment No.: 118.

Facility Operating License No. NPF-43: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: February 25, 1998 (63

FR 9598).

The March 10, 1998, supplement requested a change in the

implementation period. This information was within the scope of the

original Federal Register notice and did not change the staff's initial

proposed no significant hazards considerations determination. The

Commission's related evaluation of the amendment is contained in a

Safety Evaluation dated May 28, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

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

Michigan

Date of application for amendment: November 22, 1995 (NRC-95-0124),

as supplemented February 19, April 19, May 3, June 12, and December 4,

1996, January 30 and August 7, 1997, and April 27 and May 22, 1998.

Brief description of amendment: The amendment revises technical

specification (TS) 3.8.1.1 to change the emergency diesel generator

(EDG) allowed outage time from 3 to 7 days and add a requirement to

verify that combustion turbine-generator 11-1 is available prior to

removing an EDG from service. In addition, in accordance with draft

staff guidance for risk-informed amendments, a section is added to the

Administrative Controls Section of the TS describing the licensee's

configuration risk management program. The associated Bases are also

revised. The November 22, 1995, submittal also requested changes to the

testing and reporting requirements for the EDGs. These aspects were

addressed in Amendment No. 107 to the TS issued on June 20, 1996. The

staff's action on the licensee's request is now complete.

Date of issuance: June 2, 1998.

Effective date: June 2, 1998, with full implementation within 60

days.

Amendment No.: 119.

Facility Operating License No. NPF-43. Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: February 28, 1996 (61

FR 7550) with a supplemental notice on May 1, 1998 (63 FR 24195).

[[Page 33116]]

The February 19, April 19, May 3, June 12, and December 4, 1996,

August 7, 1997, and May 22, 1998, submittals provided clarifying

information within the scope of the Federal Register notices and did

not change the staff's initial proposed no significant hazards

considerations determinations.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 2, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

Ellis Reference and Information Center, 3700 South Custer Road, Monroe,

Michigan 48161.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver

Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

Date of application for amendments: March 17, 1998, as supplemented

May 14, 1998.

Brief description of amendments: These amendments revise Action 34

of technical specification (TS) Table 3.3-3, ``Engineered Safety

Feature Actuation System Instrumentation.'' Action 34 is applicable to

Functional Units 6.b., ``Grid Degraded Voltage (4.16 kV Bus),'' and

6.c., ``Grid Degraded Voltage (480 v Bus).'' Revised Action 34 requires

that with one degraded grid voltage monitoring channel inoperable, the

inoperable channel be placed in the tripped condition within one hour;

otherwise, immediately enter the applicable action statement(s) for the

associated emergency diesel generator made inoperable by the degraded

voltage start instrumentation. The revision to Action 34 also requires

that with two degraded grid voltage monitoring channels inoperable,

within one hour restore at least one of the channels to operable status

and place the other channel in the tripped condition; otherwise, the

associated emergency diesel generator would be declared inoperable and

its applicable action statement(s) entered. Corresponding changes have

also been made in the bases for TS 3/4.3.2 and the BVPS-2 TS Index

pages.

Date of issuance: May 27, 1998.

Effective date: Effective immediately, to be implemented within 60

days (both units).

Amendment Nos.: 214 and 91.

Facility Operating License Nos. DPR-66 and NPF-73: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: April 22, 1998 (63 FR

19969).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 27, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA 15001.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver

Valley Power Station, Unit Nos. 1 and 2, (BVPS-1 and BVPS-2)

Shippingport, Pennsylvania

Date of application for amendments: March 16, 1998, as supplemented

May 14, 1998.

Brief description of amendments: These amendments revise technical

specification (TS) Table 4.3-1 to add footnote 6 to the channel

calibration requirement for all instrument channels that are provided

with an input from neutron flux detectors. Footnote 6 provides that

neutron detectors may be excluded from channel calibrations. In

addition, BVPS-1 TS Table 4.3-1 is being revised to add channel

calibration requirements to items 2.b. (Power Range, Neutron Flux, Low

Setpoint), 5. (Intermediate Range, Neutron Flux), 6. (Source Range,

Neutron Flux (Below P-10)), and 23. (Reactor Trip System Interlocks P-

6, P-8, P-9, and P-10). Furthermore, changes are being made to correct

page numbers in the BVPS-2 TS Index and to add corresponding changes to

the TS Bases for both units.

Date of issuance: May 28, 1998.

Effective date: Both units, effective immediately, to be

implemented within 60 days.

Amendment Nos.: 215 and 92.

Facility Operating License Nos. DPR-66 and NPF-73: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: April 22, 1998 (63 FR

19969).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 28, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: B. F. Jones Memorial Library,

663 Franklin Avenue, Aliquippa, PA 15001.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey

Point Plant Units 3 and 4, Dade County, Florida

Date of application for amendments: January 9, 1998, as

supplemented by letter dated April 20, 1998.

Brief description of amendments: The amendments permit the use of

fuel with ZIRLO cladding.

Date of issuance: May 12, 1998.

Effective date: May 12, 1998.

Amendment Nos. 196 and 190.

Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: February 25, 1998 (63

FR 9605).

The April 20, 1998 letter provided clarifying information that did

not change the initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 12, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear

Generating Station, Ocean County, New Jersey

Date of application for amendment: December 10, 1997.

Brief description of amendment: The amendment clarifies sections of

the Technical Specifications that have been demonstrated to be unclear

or conflicting.

Date of Issuance: June 4, 1998.

Effective date: June 4, 1998, to be implemented within 30 days.

Amendment No.: 195.

Facility Operating License No. DPR-16: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 28, 1998 (63 FR

4313).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated June 4, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

Reference Department, 101 Washington Street, Toms River, NJ 08753.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station Unit No. 2, Oswego County, New York

Date of application for amendment: December 15, 1997, as

supplemented by letter dated April 24, 1998.

Brief description of amendment: This amendment changes Technical

Specifications 2.1.2 and 3.4.1.1 to revise the minimum critical power

ratio safety limits for fuel operating cycle 7 for two-loop and single-

loop recirculation operation.

Date of issuance: June 4, 1998.

Effective date: As of the date of issuance to be implemented before

[[Page 33117]]

startup of the Unit 2 reactor to begin fuel operating cycle 7.

Amendment No.: 82.

Facility Operating License No. DPR-63: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: January 28, 1998 (63 FR

4314).

The April 24, 1998, submittal provided clarifying information that

did not alter the initial no significant hazards consideration

determination. The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 4, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126

Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone

Nuclear Power Station, Unit No. 2, New London County, Connecticut

Date of application for amendment: September 2, 1997.

Brief description of amendment: The amendment corrects several

compliance issues as identified in Licensee Event Report 97-022-00

``Technical Specification Violations'' dated July 9, 1997, by rewording

the text; changing terminology and numbering; combining two Technical

Specifications (TSs) into one; changing the allowed outage times;

specifying guidance for entering into TS 3.0.3; changing a definition;

changing surveillance requirments, and updating the TS Bases section to

reflect changes.

Date of issuance: May 26, 1998.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 215.

Facility Operating License No. DPR-65: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 24, 1997 (62

FR 50008).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 26, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of application for amendment: October 15, 1997, as

supplemented January 23 and April 8, 1998.

Brief description of amendment: The amendment revises the action

statements and the instrumentation trip setpoint tables in the

Technical Specifications for the reactor trip system and engineered

safety feature actuation system instrumentation. In addition, the

amendment (1) decreases the reactor trip setpoint for the reactor

coolant pump low shaft speed (underspeed trip setpoint) from 95.8

percent to 92.4 percent of rated speed, (2) makes editorial changes,

and (3) changes the Bases to reflect the new methodology.

Date of issuance: May 26, 1998.

Effective date: As of the date of issuance, to be implemented

within 60 days.

Amendment No.: 159.

Facility Operating License No. NPF-49: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 19, 1997 (62

FR 61842).

The January 23 and April 8, 1998, submittals provided clarifying

and additional information that did not change the scope of the October

15, 1997, application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 26, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of application for amendment: April 7, 1998.

Brief description of amendment: The amendment replaces the

pressurizer maximum water inventory requirement with a pressurizer

maximum indicated level requirement. The amendment also makes editorial

changes and modifies the associated Bases section.

Date of issuance: May 27, 1998.

Effective date: As of the date of issuance, to be implemented

within 30 days.

Amendment No.: 160.

Facility Operating License No. NPF-49: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 23, 1998 (63 FR

20219).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 27, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,

49 Rope Ferry Road, Waterford, Connecticut.

Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone

Nuclear Power Station, Unit No. 3, New London County, Connecticut

Date of application for amendment: April 14, 1998, as supplemented

May 7, 1998, and two letters dated June 4, 1998.

Brief description of amendment: The amendment changes Technical

Specification 3/4.4.4, Relief Valves, to ensure that the automatic

capability of the power-operated relief valves (PORVs) to relieve

pressure is maintained when these valves are isolated by closure of the

block valves. The amendment also makes editorial changes, adds PORV

surveillance requirements, and modifies the associated Bases section.

Date of issuance: June 5, 1998.

Effective date: As of the date of issuance, to be implemented

within 30 days.

Amendment No.: 161.

Facility Operating License No. NPF-49: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 20, 1998 (63 FR

19532).

The May 7, 1998, letter and the two letters dated June 4, 1998,

provide clarifying information that did not change the scope of the

April 14, 1998, application and the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated June 5, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

Three Rivers Community-Technical College, 574 New London Turnpike,

Norwich, Connecticut, and the

[[Page 33118]]

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

Connecticut.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,

California

Date of application for amendments: December 23, 1997.

Brief description of amendments: The amendments changed the

combined Technical Specifications (TS) for the Diablo Canyon Power

Plant (DCPP) Unit Nos. 1 and 2 to revise TS 3/4.7.1.1, Table 3.7-1,

``Maximum Allowable Power Range Neutron Flux High Setpoint With

Inoperable Steam Line Safety Valves.'' The power range (PR) neutron

flux high setpoints were changed based on revised calculational

methodologies for 1, 2, or 3 inoperable MSSVs per steam generator (SG).

The proposed TS change lowered the PR neutron flux high setpoints when

2 or 3 MSSV are inoperable per loop such that the maximum power level

allowed would be within the heat removing capability of the remaining

operable MSSVs. Although the method for calculating the maximum power

level allowed when one MSSV per loop is inoperable was revised, the

results were not and the limit remained the same. The associated Bases

were also revised.

Date of issuance: May 28, 1998.

Effective date: May 28, 1998, to be implemented within 30 days of

issuance.

Amendment Nos.: Unit 1-125; Unit 2-123.

Facility Operating License Nos. DPR-80 and DPR-82: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: April 22, 1998 (63 FR

19975).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 28, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

University, Robert E. Kennedy Library, Government Documents and Maps

Department, San Luis Obispo, California 93407.

Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear

Generating Station, Unit No. 1, Salem County, New Jersey

Date of application for amendment: March 26, 1998.

Brief description of amendment: The amendment revises Technical

Specification 3.1.3.3, ``Rod Drop Time,'' to change the applicability

from Mode 3 (hot shutdown) to Modes 1 and 2 (startup and power

operation).

Date of issuance: June 4, 1998.

Effective date: As of date of issuance to be implemented within 60

days.

Amendment No.: 211.

Facility Operating License No. DPR-70: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 22, 1998 (63 FR

19978). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated June 4, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

Southern Nuclear Operating Company, Inc., Georgia Power Company,

Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,

City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch

Nuclear Plant, Units 1 and 2, Appling County, Georgia

Date of application for amendments: May 30, 1997, as supplemented

April 1, 1998.

Brief description of amendments: The amendments revise the

Technical Specification requirements to reflect a design modification

that changes the power sources to valves associated with the low

pressure coolant injection mode of the residual heat removal system.

Date of issuance: June 2, 1998.

Effective date: As of the date of issuance to be implemented prior

to startup from the next refueling outage for both units.

Amendment Nos.: Unit 1-211; Unit 2-152.

Facility Operating License Nos. DPR-57 and NPF-5: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: July 16, 1997 (62 FR

38139).

The April 1, 1998, submittal provided clarifying information that

did not change the scope of the May 30, 1997, application and the

initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 2, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia.

Southern Nuclear Power Company, Inc., Georgia Power Company, Oglethorpe

Power Corporation, Municipal Electric Authority of Georgia, City of

Dalton, Georgia, Docket Nos. 50-424 and 50-425, Vogtle Electric

Generating Plant, Units 1 and 2, Burke County, Georgia

Date of application for amendments: November 20, 1997, as

supplemented by letter dated April 16, 1998.

Brief description of amendments: The proposed changes to the

Technical Specifications (TS): (1) Remove the inequalities applied to

the ``Trip Setpoint'' column of TS Table 3.3.1-1, ``Reactor Trip System

Instrumentation'' and TS Table 3.3.2-1, ``Engineered Safety Feature

Actuation System Instrumentation'' and revise the ``Trip Setpoint''

column to read ``Nominal Trip Setpoint;'' (2) Add footnotes (n) and (i)

to TS Tables 3.3.1-1 and 3.3.2-1, respectively, to include criteria for

channel operability, reset, and calibration tolerance about the trip

setpoint. These footnotes also allow for the trip setpoint to be set

more conservatively than the Nominal Trip Setpoint value as necessary

in response to plant conditions; (3) The Allowable Value for TS Table

3.3.1-1, Function 14.b, Turbine Trip--Turbine Stop Valve Closure, would

be revised from ``[greater than or equal to] 96.7% open'' to ``[greater

than or equal to] 90% open;'' (4) Revise footnotes (l) and (m) of TS

Table 3.3.1-1 to refer to Nominal Trip Setpoint and delete the

inequalities applied to the trip setpoints; (5) Delete the superscript

``(a)'' from the ``Trip Setpoint'' column on page 6 of 8 of Table

3.3.1-1; (6) Revise the inequality for the Engineered Safety Feature

Actuation System Allowable Value for Steam Line Pressure--Low (Table

3.3.2-1, Function 1.e) from ``[less than or equal to]'' to ``[greater

than or equal to];'' and (7) Revise associated TS Bases to reflect the

TS revisions.

Date of issuance: June 1, 1998.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment Nos.: Unit 1-101; Unit 2-79.

Facility Operating License Nos. NPF-68 and NPF-81: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 31, 1997 (62

FR 68318).

The supplement dated April 16, 1998, provided clarifying

information that did not change the scope of the November 20, 1997,

application and the initial proposed no significant hazards

determination.

[[Page 33119]]

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated June 1, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Burke County Library, 412

Fourth Street, Waynesboro, Georgia.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam

Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

Date of amendment request: May 1, 1995 (TXX-95090).

Brief description of amendments: These amendments revise section 3/

4.8.1 of the Technical Specifications (TSs) to reduce the minimum fuel

oil volume requirement during MODES 5 and 6 for an operable emergency

diesel generator (EDG) and allow continued OPERABLE status of diesel

generators during all MODES for 48 hours with greater than a 6 day

supply of diesel fuel for a given EDG.

Date of issuance: May 22, 1998.

Effective date: May 22, 1998, to be implemented within 30 days.

Amendment Nos.: Unit 1--Amendment No. 60; Unit 2--Amendment No. 46.

Facility Operating License Nos. NPF-87 and NPF-89: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 21, 1995 (60 FR

32373).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 22, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Texas at

Arlington Library, Government Publications/Maps, 702 College, P.O. Box

19497, Arlington, TX 76019.

Washington Public Power Supply System, Docket No. 50-397, Nuclear

Project No. 2, Benton County, Washington

Date of application for amendment: December 4, 1997, as

supplemented by letters dated January 28, 1998, March 3, 1998, March 9,

1998, and April 24, 1998.

Brief description of amendment: The amendment permits the continued

used of the existing Siemens Power Corporation minimum critical power

ratio (MCPR) safety limits for WNP-2 Fuel Cycle 14 and changes the ASEA

Brown Boveri (ABB) MCPR safety limit for single loop operation from

1.08 for Cycle 13 to 1.09 for Cycle 14.

Date of issuance: May 29, 1998.

Effective date: May 29, 1998, to be implemented within 30 days from

the date of issuance.

Amendment No.: 154.

Facility Operating License No. NPF-21: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: January 14, 1998 (63 FR

2284).

The January 28, 1998, March 3, 1998, March 9, 1998, and April 24,

1998, supplemental letters provided additional clarifying information

and did not change the original no significant hazards consideration.

The Commission's related evaluation of the amendment is contained in a

Safety Evaluation dated May 29, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Richland Public Library, 955

Northgate Street, Richland, Washington 99352.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: February 25, 1998.

Brief description of amendment: The amendment revises the Technical

Specifications to implement performance-based containment leakage

testing under Option B of 10 CFR 50, Appendix J.

Date of issuance: May 28, 1998.

Effective date: May 28, 1998.

Amendment No.: 136.

Facility Operating License No. DPR-43: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 8, 1998 (63 FR

17237).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 28, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of amendment request: May 8, 1998, as supplemented by letter

dated May 11, 1998.

Brief description of amendment: The amendment adds a new Action

Statement to Technical Specification 3/4.3.2, Table 3.3-3, Functional

Unit 7.b., Refueling Water Storage Tank Level--Low-Low Coincident With

Safety Injection.

Date of issuance: May 28, 1998.

Effective date: May 28, 1998.

Amendment No.: 117.

Facility Operating License No. NPF-42: The amendment revised the

Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration: Yes (63 FR 26829 dated May 14, 1998). The notice

provided an opportunity to submit comments on the Commission's proposed

no significant hazards consideration determination. No comments have

been received. The notice also provided for an opportunity to request a

hearing by June 15, 1998, but indicated that if the Commission makes a

final no significant hazards consideration determination any such

hearing would take place after issuance of the amendment. The

Commission's related evaluation of the amendment, finding of exigent

circumstances, consultation with the State of Kansas and final

determination of no significant hazards consideration are contained in

a Safety Evaluation dated May 28, 1998.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621.

Attorney for Licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.

NRC Project Director: William H. Bateman.

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

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects--III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 98-16012 Filed 6-16-98; 8:45 am]

BILLING CODE 7590-01-P