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 March 1, 1999, through March 12, 1999. The

last biweekly notice was published on March 10, 1999 (64 FR 11958).

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 April 23, 1999, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

who has filed a petition for

[[Page 14279]]

leave to intervene or who has been admitted as a party may amend the

petition without requesting leave of the Board up to 15 days prior to

the first prehearing conference scheduled in the proceeding, but such

an amended petition must satisfy the specificity requirements described

above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

issue of law or fact. Contentions shall be limited to matters within

the scope of the amendment under consideration. The contention must be

one which, if proven, would entitle the petitioner to relief. A

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

Office of the General Counsel, U.S. Nuclear Regulatory Commission,

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for 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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-

529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.

1, 2, and 3, Maricopa County, Arizona

Date of amendments request: December 16, 1998.

Description of amendments request: The proposed amendment would

revise Technical Specification (TS) 3.8.1, ``AC Sources--Operating,''

and TS 3.3.7, ``Diesel Generator (DG)--Loss of Voltage Start (LOVS).''

The proposed amendment will (1) change Condition G of TS 3.8.1 to

ensure that the appropriate actions will be taken to prevent double

sequencing of safety-related loads, and (2) change TS 3.3.7 to ensure

that the setpoint allowable values for the degraded voltage and the

loss of voltage relays reflect the required function of the relays.

Basis for proposed no significant hazards consideration determination:

As required by 10 CFR 50.91(a), the licensee has provided its analysis

of the issue of no significant hazards consideration, which is

presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed amendment will change Condition G of Technical

Specification 3.8.1. These changes will ensure that the appropriate

actions will be taken to prevent double sequencing of safety-related

loads. This change is required to assure the capability of the

offsite circuits ``to effect a safe shutdown and to mitigate the

effects of an accident'' in accordance with Regulatory Guide 1.93.

The proposed amendment will also change the setpoint allowable

values for the degraded voltage and the loss of voltage relays in

Technical Specification Surveillance Requirement (SR) 3.3.7.3. The

proposed changes do not involve any physical changes to plant

equipment. The actions required by the TS amendment will identify

when an offsite circuit does not meet its required capability and

provides actions to restore the required capability. The proposed

changes are intended to identify and correct the conditions (voltage

and/or loading) required to prevent the possibility of a double

sequencing event. Therefore, this change ensures that power will be

supplied to the ESF [engineered safety feature] loads following a

loss of offsite power event described in UFSAR [Updated Final Safety

Analysis Report] 15.2.6.1. For other events discussed in the UFSAR,

the electrical distribution system is an event mitigator. This

change will ensure that the electrical distribution system will

continue to meet this requirement. The proposed changes will not

effect the function of the DG loss of voltage start as required by

the design basis and safety analysis. Therefore, the proposed change

does not involve a significant increase in the probability of an

accident previously evaluated.

The proposed changes do not involve any physical changes to

plant equipment. The proposed changes ensure that appropriate

controls are in place to prevent a double sequencing event. The

proposed changes consider the factors in preventing a double

sequencing event such as pretrip voltage, load, number of units on

line, and number of transmission lines in service. These are factors

which could affect post trip voltage. The actions associated with

this change will identify and mitigate the condition where an

offsite circuit does not meet its required capability and, as such,

do not result in new or revised accident sequences. The proposed

changes will not effect the function of the DG loss of voltage start

as required by the design basis and safety analysis. Therefore, the

proposed change does 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.

The proposed amendment will change Condition G of Technical

Specification 3.8.1. These changes will ensure that the appropriate

actions will be taken to prevent double sequencing of safety-related

loads. This change is required to assure the capability of the

offsite circuits ``to effect a safe shutdown and to mitigate the

effects of an accident'' in accordance with Regulatory

[[Page 14280]]

Guide 1.93. The proposed amendment will also clarify the setpoint

allowable values for the degraded voltage and the loss of voltage

relays in Technical Specification Surveillance Requirement (SR)

3.3.7.3. The proposed changes do not change the operation of any

system or equipment, nor do they create a new type of malfunction.

The proposed changes prevent double sequencing and do not create the

possibility of any other malfunction. The actions associated with

this change will identify and mitigate the condition where an

offsite circuit does not meet its required capability. The proposed

changes will not effect the function of the DG loss of voltage start

as required by the design basis and safety analysis. Therefore, the

proposed changes do not create the possibility of a new or different

kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed amendment will change Condition G of Technical

Specification 3.8.1. These changes will ensure that the appropriate

actions will be taken to prevent double sequencing of safety-related

loads. This change is required to assure the capability of the

offsite circuits ``to effect a safe shutdown and to mitigate the

effects of an accident'' in accordance with Regulatory Guide 1.93.

The proposed amendment will also change the setpoint allowable

values for the degraded voltage and the loss of voltage relays in

Technical Specification Surveillance Requirement (SR) 3.3.7.3. The

proposed changes ensure that the units will be in conformance with

GDC 17, Electric Power Systems (basis for TS 3.8.1). The required

actions of the proposed change will ensure that the single failure

analyses and safety analysis are maintained. The actions associated

with this change will identify and mitigate the condition where an

offsite circuit does not meet its required capability. The proposed

changes ensure that the bases for the current TS are maintained. The

proposed changes will not effect the function of the DG loss of

voltage start as required by the design basis and safety analysis.

Therefore, the proposed changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendments request involve no significant hazards consideration.

Local Public Document Room location: Phoenix Public Library, 1221

N. Central Avenue, Phoenix, Arizona 85004.

Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary

and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail

Station 9068, Phoenix, Arizona 85072-3999.

NRC Project Director: William H. Bateman.

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 amendment request: February 26, 1999.

Description of amendment request: The proposed amendment would

revise the Table Notations for Technical Specification (TS) Table 3.3-

4, ``Engineered Safety Features Actuation System Instrumentation Trip

Setpoints.'' Specifically, the time constants used in the lead-lag

controller for Steam Line Pressure--Low (Table item 1.e.) are

t1 greater than or equal to 50 seconds and t2

greater than or equal to 5 seconds. The proposed amendment would revise

t2 to less than or equal to 5 seconds. Also, the time

constant used in the rate-lag controller for Negative Steam Line

Pressure Rate--High (Table item 4.e.) is less than or equal to 50

seconds. The proposed amendment would revise this time constant to

greater than or equal to 50 seconds.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Correcting the time constants will ensure conservative

calibration of the Engineered Safety Feature Actuation System

instrumentation. The proposed amendment will not introduce any new

equipment or require existing equipment to function different from

that previously evaluated in the Final Safety Analysis Report (FSAR)

or TS. Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

2. The proposed amendment does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

Correcting the time constants will ensure conservative

calibration of the Engineered Safety Feature Actuation System

instrumentation. The proposed amendment will not introduce any new

equipment or require existing equipment to function different from

that previously evaluated in the Final Safety Analysis Report (FSAR)

or TS. The proposed amendment will not create any new accident

scenarios, because the change does not introduce any new single

failures, adverse equipment or material interactions, or release

paths. Therefore, the proposed change does not create the

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

previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

Correcting the time constants will ensure conservative

calibration of the Engineered Safety Feature Actuation System

instrumentation. Therefore, the proposed change does not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

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

Attorney for licensee: William D. Johnson, Vice President and

Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,

Raleigh, North Carolina 27602.

NRC Project Director: Cecil Thomas.

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

County, Michigan

Date of application for amendment: March 26, 1997.

Brief description of amendment: The amendment revises modifies

Technical Specification sections 3.6 and 4.5 by removing the list of

containment isolation valves in accordance with Generic Letter 91-08,

``Removal of Components Lists from Technical Specifications,'' dated

May 6, 1991, and by revising requirements related to containment

pressure and containment temperature. Additionally, several editorial

changes are made to emulate the format and content of NUREG-1432,

``Standard Technical Specifications, Combustion Engineering Plants.''

Date of issuance: February 22, 1999.

Effective date: February 22, 1999.

Amendment No.: 184.

Facility Operating License No. DPR-20. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 17, 1997 (62

FR 66136)

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 22, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

[[Page 14281]]

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

County, Michigan

Date of amendment request: September 3, 1997.

Description of amendment request: The proposed amendment includes

the following changes to the station technical specification (TS):

(a) TS Action Statement 3.14a is replaced by a revised condition

description for TS Action Statement 3.17.1.6 in the instrumentation

systems section. Also, the maximum control room temperature at which a

shutdown must be initiated is revised from 120 deg.F [degrees

Fahrenheit] to 90 deg.F, and a time limit for reaching the hot

shutdown condition is specified;

(b) TS 3.14b is replaced with two limiting conditions for operation

(LCOs), 3.14.1 and 3.14.2, addressing, respectively, the filtration and

cooling functions of the CRHVAC [control room heating, ventilation, and

air conditioning] system. These proposed LCOs emulate the standard TS

(NUREG 1432) for control room ventilation;

(c) TS Table 4.2.3 surveillance requirement (SR) number 3,

verification of control room temperature, is moved to SR Table 4.17.1,

for the reactor protection system (RPS); and

(d) other administrative changes.

The licensee classified each change as either administrative or

more restrictive. An administrative change is editorial in nature,

involves only movement of requirements within the TS without affecting

their technical content, or clarifies existing TS requirements. A more

restrictive change adds new requirements, or revises existing

requirements resulting in more conservative or additional operational

restrictions.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

against the standards of 10 CFR 50.92(c). The NRC staff's review is

presented below:

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

probability or consequences of an accident previously evaluated?

The proposed changes to TS 3.14a and TS 3.14b constitute either

new, or more restrictive requirements that provide additional assurance

that equipment conforms to the plant design basis and will operate

reliably when called upon. These changes represent additional

restrictions on plant operation that enhance safety and are consistent

with the standard TS. The proposed change to TS Table 4.2.3 of moving

SR item number 3 to TS Table 4.17.1, and other administrative changes

are editorial in nature or involve the reorganization or reformatting

of TS requirements without affecting technical content or operational

restrictions. The proposed changes do not result in any substantive

change in operating requirements or the intent of these requirements,

and are consistent with the Commission's regulations. Therefore, these

changes cannot involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

different kind of accident from any previously evaluated?

The proposed changes to TS 3.14a and TS 3.14b constitute either

new, or more restrictive requirements that provide additional assurance

that equipment conforms to the plant design basis and will operate

reliably when called upon. These changes represent additional

restrictions on plant operation that enhance safety and are consistent

with the standard TS. The proposed change to TS Table 4.2.3 of moving

SR number 3 to TS Table 4.17.1, and other administrative changes are

editorial in nature or involve the reorganization or reformatting of TS

requirements without affecting technical content or operational

restrictions. The proposed changes do not result in any substantive

change in operating requirements or the intent of these requirements,

and are consistent with the Commission's regulations. Therefore, these

changes cannot create the possibility of a new or different kind of

accident from any previously evaluated.

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

margin of safety?

The proposed changes to TS 3.14a and TS 3.14b constitute either

new, or more restrictive requirements that provide additional assurance

that equipment conforms to the plant design basis and will operate

reliably when called upon. These changes represent additional

restrictions on plant operation that enhance safety and are consistent

with the standard TS. The proposed change to TS Table 4.2.3 of moving

SR number 3 to TS Table 4.17.1, and other administrative changes are

editorial in nature or involve the reorganization or reformatting of TS

requirements without affecting technical content or operational

restrictions. The proposed changes do not result in any substantive

change in operating requirements or the intent of these requirements,

and are consistent with the Commission's regulations. Therefore, these

changes cannot involve a significant reduction in a margin of safety.

Based on this review, it appears that the three standards of 10 CFR

50.92(c) are satisfied. Therefore, the NRC staff proposes to determine

that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423-3698.

Attorney for licensee: Arunas T. Udrys, Esquire, Consumers Energy

Company, 212 West Michigan Avenue, Jackson, Michigan 49201.

NRC Project Director: Cynthia A. Carpenter.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: March 1, 1999.

Description of amendment request: The proposed amendments would

revise Oconee Nuclear Station, Units 1, 2, and 3 Improved Technical

Specification (ITS) 3.3.8 to only require two channels for the reactor

coolant system hot leg temperature function. The current TSs require

two channels per loop. This requirement was incorrectly specified

during the ITS conversion.

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 modifies ITS Table 3.3.8-1 to only require

two channels for RCS [Reactor Coolant System] Hot Leg Temperature

Function. These instruments provide indication only and are not

considered as initiators of any analyzed event. The proposed change

does not involve a physical alteration of the plant. No new or

different equipment is being installed, and no installed equipment

is being operated in a new or different manner. No set points for

parameters which initiate protective or mitigative action are being

changed. Therefore, the proposed changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

from any kind of accident previously evaluated:

The proposed change does not involve a physical alteration of

the plant. No new or different equipment is being installed, and no

[[Page 14282]]

installed equipment is being operated in a new or different manner.

No set points for parameters which initiate protective or mitigative

action are being changed. As a result, no new failure modes are

being introduced. Therefore, this proposed amendment will not create

the possibility of any new or different kind of accident.

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

The margin of safety for PAM [post accident monitoring]

instrumentation is based on the availability and capability of the

instrumentation to provide the required operator information. The

proposed change maintains requirements within the safety analyses

and licensing basis and has no effect on the availability and

capability of the PAM function. 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: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

Attorney for licensee: Ann W. Cottington, Winston and Strawn, 1200

17th Street, NW., Washington, DC.

NRC Project Director: Herbert N. Berkow.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of amendment request: March 1, 1999.

Description of amendment request: The proposed amendments to

Improved Technical Specification (ITS) 3.9, ``Refueling Operations,''

Subsection 3.9.3, ``Containment Penetrations,'' Limiting Condition for

Operation 3.9.3.b would add a Note to state that the emergency air lock

door is not required to be closed when it is sealed with a temporary

cover plate. The temporary cover plate contains penetrations that are

used for such refueling outage services as cables, pneumatic tubing,

and hoses.

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:

This proposed change has been evaluated against the standards in

10 CFR 50.92 and has been determined to involve no significant

hazards, in 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?

The proposed change allows the use of a temporary cover plate as

a seal for the emergency air lock during refueling operations in

lieu of an air lock door. Duke [Duke Energy Corporation] analyses

for Oconee Nuclear Station (ONS) does not credit containment

closure. Therefore, use of the temporary cover plate does not affect

offsite doses, which were previously calculated to be well within 10

CFR 100 limits. As such, the proposed change does not involve a

significant increase in the probability or consequences of an

accident.

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

from any accident previously evaluated.

The fuel handling accident inside containment analyses discussed

in the Updated Final Safety Analysis Report section 15.11 bound the

proposed change. No new or different type of accident will occur

because of the temporary cover plate placement.

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

Placing the temporary cover plate in the emergency air lock will

still meet the intent of containment closure. The building pressure

does not increase during a fuel handling accident and fission

products will be contained. The fuel handling accident inside

containment analyses does not credit containment closure for

reducing offsite dose. As such, the proposed change does not involve

a significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200

17th Street, NW., Washington, DC.

NRC Project Director: Herbert N. Berkow.

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

Point Plant Units 3 and 4, Dade County, Florida

Date of amendment request: February 24, 1999.

Description of amendment request: The proposed amendments would

change Technical Specification (TS)

3/4.7.4 to remove the restriction to monitor the Ultimate Heat Sink

(UHS) temperature only in the Intake Cooling Water (ICW) bay and prior

to the ICW pumps. This change would permit the option of monitoring the

UHS temperature after the ICW pumps but prior to the component cooling

water heat exchangers, which is considered to be equivalent to

temperature monitoring before the ICW pumps.

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 method of monitoring the Ultimate Heat Sink temperature is

not considered in, and has no effect on, the probability of any type

of accident initiating sequence. The proposed changes will permit

other means of monitoring the Ultimate Heat Sink that have been

evaluated to be equivalent to the current method permitted. As the

monitoring will continue to be performed by equal means, the

consequences of any accident previously evaluated will not be

affected.

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 change will permit other means of monitoring the

Ultimate Heat Sink temperature, which will be equal to the methods

currently employed. The continued monitoring of this variable by

equivalent means cannot create the possibility of a new or different

type of accident.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety.

The Ultimate Heat Sink temperature is an input assumption used

in the accident analysis and in evaluation of component design. This

temperature limit is not being altered by this change, only the

permissible means of monitoring this variable. As any new methods

employed are expected to be equivalent to those currently used, no

reduction in any margin of safety will result.

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: Florida International

University, University Park, Miami, Florida 33199.

Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light,

P.O. Box 14000, Juno Beach, Florida 33408-0420.

NRC Project Director: Cecil O. Thomas.

[[Page 14283]]

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: February 2, 1999.

Description of amendment request: The proposed amendment revises

the Technical Specifications (TS) to expand the scope of systems and

test requirements considered under TS 4.5.4 ``Engineered Safeguards

Feature (ESF) Systems Leakage,'' and increases the maximum allowable

leakage for those portions of the ESF system outside containment. The

proposed amendment also includes revised the Bases for TS 3.15.3,

``Auxiliary and Fuel Handling Building Air Treatment System,'' to

clarify system design requirements and accident analysis

considerations.

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. No

physical modifications which would change structures, systems or

components are proposed by this TSCR [technical specification change

request] for surveillance changes in Technical Specification 4.5.4

and its Bases. The proposed increase in the ESF Systems leakage rate

acceptance limit has no effect on the performance of ESF systems

during a DBA [design basis accident]. The proposed changes are

supported by a revised MHA [maximum hypothetical accident] dose

calculation using updated X/Q values and calculation assumptions.

The MHA dose consequence analysis yields dose results that are below

the 10 CFR 100 guidelines for both the EAB [exclusion area boundary]

and LPZ [low population zone]. The calculated Control Room

Habitability Evaluation does not exceed the permissible annual

occupational exposure limit of 50 Rem to the thyroid as specified in

10 CFR 20.1201(a)(ii). In addition, the potential thyroid exposure

can be mitigated by the availability of self-contained breathing

apparatus and potassium iodide. Therefore, the changes would not

involve a significant increase in the consequences of accidents

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 previously evaluated. This TSCR does not

involve any physical modifications that would affect structures,

systems, or components, nor does it involve any changes in plant

operation.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety. This TSCR does not involve changes to the Technical

Specification defined Safety Limits, Limiting Conditions for

Operation, and does not involve any change to safety system

setpoints for operation. Therefore, the proposed changes do not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Law/Government Publications

Section, State Library of Pennsylvania, (Regional Depository) Walnut

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105

Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,

Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.

NRC Project Director: Elinor G. Adensam.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear

Generating Plant, Wright County, Minnesota

Date of amendment request: February 12, 1999.

Description of amendment request: The proposed amendment would

change the Technical Specifications (TS) to (1) allow reactor vessel

hydrostatic and leakage tests without maintaining primary containment

integrity, (2) establish a limit and a surveillance requirement on

reactor coolant activity when reactor coolant temperature is above

212 deg.F, the reactor is not critical, and primary containment has not

been established, and (3) correct a punctuation error.

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 amendment will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes do not increase the probability of an

accident since reactor vessel hydrostatic and leakage tests would be

performed with the reactor vessel nearly water solid, at nominal

operating pressure, not critical and at low decay heat values which

minimizes the energy stored in the reactor vessel. Under this

proposed change a limit on reactor coolant activity is established

that provides adequate assurance that the consequences of a large

primary system break during reactor vessel hydrostatic and leakage

test conditions will be conservatively bounded by the consequences

of a postulated main steam line break outside of primary

containment. Low pressure emergency core cooling systems are

required to be operable during reactor vessel hydrostatic and

leakage test providing assurance that adequate core cooling can be

achieved to preclude fuel failures and subsequent increases in

reactor coolant activity in the event of a large primary system

break. The reduced stored energy in the reactor vessel and proposed

limit on reactor coolant activity ensures there is no increase in

the probability or consequences of an accident previously evaluated.

The proposed amendment will not create the possibility of a new

or different kind of accident from any accident previously analyzed.

The proposed changes do not introduce any new accident

initiators or failure mechanisms since the changes do not involve

any changes to the structures, systems, or components. They also do

not involve any change to the operation of systems, and alter

procedures only to the extent that 212 deg.F may be exceeded during

reactor vessel hydrostatic and leakage testing without maintaining

primary containment integrity. Without maintaining primary

containment integrity, a large primary system break during a reactor

vessel hydrostatic or leakage test would result in the same kind of

accident as would a main steam line break outside primary

containment during normal operation. Therefore, the proposed TS

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

accident, from any accident previously evaluated.

The proposed amendment will not involve a significant reduction

in the margin of safety.

Since reactor vessel hydrostatic and leakage tests are performed

nearly water solid, at nominal operating pressure, not critical and

at low decay heat values, the stored energy in the reactor vessel

during testing will be low. Under these conditions, the potential

for failed fuel and a subsequent increase in coolant activity is

minimized. Therefore, the proposed Technical Specification change

does not involve a significant reduction in the margin of safety.

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

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

satisfied. In addition, correction to the punctuation error is strictly

a grammatical change and has no effect on the three standards of 10 CFR

50.92(c). Therefore, the NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts

and

[[Page 14284]]

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

NRC Project Director: Cynthia A. Carpenter.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,

Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New

Jersey

Date of amendment request: February 8, 1999.

Description of amendment request: The proposed amendments would

revise Technical Specification 4.5.3.2.b to allow the option of using

closed and disabled automatic valves to provide the necessary isolation

function when performing safety injection and charging pump testing in

Modes 4, 5, and 6 (hot shutdown, cold shutdown, and refueling) for low

temperature over pressurization protection.

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 not involve a significant increase in the probability or

consequences of an accident previously evaluated.

In Mode 4 with the RCS [reactor coolant system] coolant

temperature less than 312 deg.F or in Modes 5 and 6 there is a

potential risk of low temperature overpressurization. Mass additions

of coolant by the safety injection and charging pumps could cause

such an event to the extent that these pump flows exceed the ability

of a single over pressure protection relief valve to protect the

system. In order to eliminate this potentiality provisions are made

to allow a maximum of one pump to be in service with the other pumps

disabled except for testing. Further provisions are made to assure

that a pump being tested can not inject into the vessel. The

proposed change merely adds an alternate method of providing this

assurance in addition to that currently provided by closing the

manual discharge valves. The proposed change offers an equivalent

means of affording the required protection.

Based upon the above, the proposed change will not increase the

probability or consequences of an accident previously analyzed.

2. Will not create the possibility of a new or different kind of

accident from any previously evaluated.

The proposed changes do not require any change in the operation

of the plant. A minor configuration change is involved in that a[ ]

disabled automatic valve in the flow path will be used in lieu of

the manual valve to provide protection. Specifically, no new

hardware is being added to the plant as part of the proposed change,

no existing equipment is being modified, and no significant changes

in operations are being introduced. Therefore, these changes will

not create the possibility of a new or different kind of accident

from any accident previously evaluated.

3. Will not involve a significant reduction in a margin of

safety.

The proposed change will not alter any assumptions, initial

conditions, or results of any accident analyses. The proposed change

maintains the level of protection. The change will, therefore, 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: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

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

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

NRC Project Director: Elinor G. Adensam.

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

Nuclear Power Plant, Wayne County, New York

Date of amendment request: March 1, 1999.

Description of amendment request: The proposed amendment would

revise the Ginna Station Improved Technical Specifications battery cell

parameters limit for specific gravity Surveillance Requirement (SR)

3.8.6.3 and SR 3.8.6.6.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Operation of Ginna Station in accordance with the proposed

change does not involve a significant increase in the probability or

consequences of an accident previously evaluated. The change is only

to correct an error in the determination of the minimum limiting

value for specific gravity of the station batteries. This does not

increase the probability of an accident previously evaluated since

the battery specific gravity is only a measure of the state of

charge of the battery and the batteries themselves are not an

accident initiator. The proposed minimum value for specific gravity,

based on the NUREG-1431 guidance, gives a higher assurance that the

battery has sufficient capacity. Therefore, the probability or

consequences of an accident previously evaluated is not

significantly increased.

(2) Operation of Ginna Station in accordance with the proposed

change 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 of the plant (i.e. no new or

different type of equipment will be added) or changes in the methods

governing normal plant operation. The change only involves

implementing a more conservative minimum limiting value for the

battery cell parameter of specific gravity. Therefore, the

possibility for a new or different kind of accident from any

accident previously evaluated is not created.

(3) Operation of Ginna Station in accordance with the proposed

change does not involve a significant reduction in a margin of

safety. The proposed change only corrects an error in the

determination of the limiting value for specific gravity. The error

is being corrected by using a more conservative value as determined

by the guidance of NUREG-1431. Therefore, this change does not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Rochester Public Library, 115

South Avenue, Rochester, New York 14610.

Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400

L Street, NW., Washington, DC 20005.

NRC Project Director: S. Singh Bajwa.

South Carolina Electric & Gas Company (SCE&G), South Carolina Public

Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear Station,

Unit No. 1, Fairfield County, South Carolina

Date of amendment request: February 18, 1999.

Description of amendment request: The proposed amendment would

revise Virgil C. Summer Nuclear Station (VCSNS) Technical Specification

(TS) 3/4.4.9 Reactor Coolant System Pressure/Temperature Limits to

incorporate the new Pressure/Temperature (PT) Limits curves consistent

with reactor vessel specimen analysis results. Additionally, the

proposed amendment would revise the Pressure/Temperature Limits Bases

section to accurately reflect current industry standards and

regulations.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes revise the Pressure/Temperature Limits

Curves to provide curves that reflect the results of the analysis

performed on reactor vessel surveillance

[[Page 14285]]

specimen W. This analysis was performed using NRC approved

methodology as documented in WCAP 14040-NP-A, dated January, 1996.

These curves provide the limits for operation of the Reactor Coolant

System during heat up, cool down, criticality, and hydrotesting. The

limits protect the reactor vessel from brittle fracture by

separating the region of acceptable operation from the region where

brittle fracture is postulated to occur. Failure of the reactor

vessel is not a VCSNS design basis accident, and, in general,

reactor vessel failure has a low probability of occurrence and is

not considered in the safety analysis.

Therefore, the change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes revise the Pressure/Temperature Limits

Curves, Section 3/4.4.9, to incorporate the results of the analysis

performed on reactor vessel specimen W. There are no plant design

changes or significant changes in any operating procedures. This

change adjusts the heat up and cool down curves to reflect the shift

in nil-ductility reference temperature of the reactor vessel as a

result of neutron embrittlement. Therefore, the change does not

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

any accident previously evaluated.

3. Does this change involve a significant reduction in margin of

safety? The proposed changes revise the Pressure/Temperature Limits

Curves, Section 3/4.4.9, to incorporate the results of the analysis

performed on reactor vessel specimen W. The new PT curves ensure

that the 10 CFR 50 Appendix G, requirements are not exceeded during

normal operation including Reactor Coolant System transients during

heat up, cool down, criticality, and hydrotesting. The new PT curves

were prepared, using approved NRC methodology, for a projected

reactor vessel neutron exposure of 32 EFPY [effective full power

years].

The new curves shift to more conservative operating limitations,

thus providing increased margin against non-ductile fractures. Since

administrative limits remain in place to ensure that 10 CFR 50

Appendix G limits are not challenged, the margin of safety described

in the TS Bases is not reduced by the proposed change. 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: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180.

Attorney for licensee: Randolph R. Mahan, South Carolina Electric &

Gas Company, Post Office Box 764, Columbia, South Carolina 29218.

NRC Project Director: Herbert N. Berkow.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-

362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San

Diego County, California

Date of amendment requests: May 8, 1996, as supplemented by letter

dated January 13, 1999.

Description of amendment requests: The January 13, 1999,

supplemental letter added an additional change to the technical

specifications (TS) to incorporate an additional restriction to the

time required to close containment when reactor coolant system (RCS)

water level is reduced during a refueling outage. This additional

restriction adds a limitation that containment must be able to be

closed within the calculated time to boil, if it is less than the

current four hour requirement. The January 13, 1999, letter supplements

the staff's proposed no significant hazards consideration determination

evaluation that was published on September 11, 1996 (61 FR 47978).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The licensee's analysis of the issue of no significant

hazards consideration on the supplemental change is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Item 6 conservatively restricts the completion time to ensure

containment closure is achieved prior to the water in the cavity

boiling, in the event of a Loss of Shutdown Cooling. This

restriction is already a self imposed requirement at San Onofre

Units 2 and 3. Incorporating it in the Technical Specification only

serves to highlight the importance of this requirement.

This change captures all periods of time when the time to boil

following a Loss of Shutdown Cooling is less than 4 hours. Having

this requirement cannot initiate an accident. However, this

requirement reduces the consequences of a Loss of Shutdown Cooling

Accident when the time to boil is less than 4 hours.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

Item 6 conservatively restricts the completion time to ensure

containment closure is achieved prior to the water in the cavity

boiling, in the event of a Loss of Shutdown Cooling. This

restriction is already a self imposed requirement at San Onofre

Units 2 and 3. Incorporating it in the Technical Specification only

serves to highlight the importance of this requirement.

This restriction cannot initiate an accident.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Item 6 conservatively restricts the completion time to ensure

containment closure is achieved prior to the water in the cavity

boiling, in the event of a Loss of Shutdown Cooling. This

restriction is already a self imposed requirement at San Onofre

Units 2 and 3. Incorporating it in the Technical Specification only

serves to highlight the importance of this requirement.

This change increases the margin of safety provided by the

Technical Specification by specifying that the containment must be

closed within 4 hours or within the calculated time to boil,

whichever is less. This change revises the Technical Specification

to specifically recognize the importance of ensuring containment

closure is achieved prior to boiling in the reactor vessel, upon a

loss of shutdown cooling.

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 requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

California, P.O. Box 19557, Irvine, California 92713.

Attorney for licensee: Douglas K. Porter, Esquire, Southern

California Edison Company, P. O. Box 800, Rosemead, California 91770.

NRC Project Director: William H. Bateman.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-

362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San

Diego County, California

Date of amendment requests: December 22, 1998.

Description of amendment requests: The proposed amendment would

modify the technical specifications (TS) to add a reference to allow

use of Westinghouse laser-welded steam generator (SG) tube sleeving.

The proposed amendment also provides typographical and editorial

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. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Steam generator tubes, tube plugging, and tube failures are

considered in the analysis of

[[Page 14286]]

accidents in the Updated Final Safety Analysis Report (UFSAR). The

steam generator tube rupture accident analysis considered the

failure of a steam generator tube. Also, inadvertent opening of a

steam generator dump valve (IOSGDV), loss of condenser vacuum

(LOCV), loss of coolant accidents (LOCAs), and feed water line break

(FWLB) accident analyses carry assumptions regarding steam generator

tube plugging. In each case, the addition of steam generator tube

sleeves to repair defective tubes will not change the probability or

consequences of any accident previously evaluated.

The sleeve configurations have been designed, analyzed, and

tested in accordance with the American Society of Mechanical

Engineers (ASME) code requirements, and mechanical testing has shown

that the sleeve and sleeve joints provide margin above acceptance

limits. Ultrasonic testing (UT) and eddy current testing (ECT) are

used to verify the adequacy of welds. Tests have demonstrated that

tube collapse will not occur due to postulated LOCA loadings.

The probability or consequences of any accident previously

evaluated is not increased because any leakage through the sleeve

assembly is fully bounded by the existing steam generator tube

rupture analysis included in the San Onofre Unit 2 and 3 Updated

Final Safety Analysis Report. Additionally, any reactor coolant flow

restriction from sleeving is addressed by a ratio of number of

sleeved tubes to be equal to a plugged tube.

Therefore, the proposed sleeving repair process will not involve

an increase in the probability or consequences of any previously

evaluated accident.

2. The proposed change does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The sleeves are captured within the steam generator tubes by

hard rolling and welding and as such are not able to physically

affect other parts of the system. The failure of a sleeve is

identical to the failure of the parent tube which has been

previously analyzed.

The use of a sleeve to span the area of degradation of the steam

generator tube restores the structural and leakage integrity of the

tubing to meet the original design requirements. Structural analysis

of the sleeve assembly shows that the requirements of the ASME code

are met. Mechanical testing has demonstrated that margin exists

above the original tube design criteria. Any hypothetical accident

as a result of any degradation in a sleeved tube would be bounded by

the existing steam generator tube rupture accident analysis.

Therefore, operation of the facility in accordance with proposed

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

of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The use of sleeves to repair degraded steam generator tubing

will maintain the integrity of the tube bundle commensurate with the

ASME Code and draft Regulatory Guide (RG) 1.121 margin requirements

for original tubing. Sleeves are components which are part of the

reactor coolant pressure boundary and meet the requirements for

Class 1 components in Section III of the ASME Boiler and Pressure

Vessel Code. The primary to secondary pressure boundary will be

maintained to the same margins as the original tubes under normal

and postulated accident conditions. The safety margins used in the

verification of the strength of the sleeve assembly are consistent

with the safety factors in the ASME Boiler and Pressure Vessel Code

used in steam generator design. Further, a test program has been

conducted by Westinghouse which demonstrated the integrity of the

lower hard rolled joint design and its capability to withstand the

design loads.

Therefore, operation of the facility with the proposed changes

will not involve a significant reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

California, P.O. Box 19557, Irvine, California 92713.

Attorney for licensee: Douglas K. Porter, Esquire, Southern

California Edison Company, P. O. Box 800, Rosemead, California 91770.

NRC Project Director: William H. Bateman.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: February 26, 1999 (TS 98-08).

Brief description of amendments: The proposed amendments would

change the Sequoyah (SQN) Technical Specifications (TS) by relocating

TS 3.7.6, ``Flood Protection Plan,'' and the associated bases to the

SQN Technical Requirements Manual (TRM). This change does not alter the

current requirements for implementation or surveillance testing of the

Flood Protection Plan and future revisions of this plan will require an

evaluation in accordance with 10 CFR 50.59.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), Tennessee Valley

Authority (TVA), 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 revision to the TS relocates the requirements for

SQN flood protection without changing the current requirements. This

administrative relocation of the requirements will not increase the

possibility of an accident.

The capability of the Flood Protection Plan will continue to

provide the same function. Changes to the relocated requirements

will be processed, in accordance with 10 CFR 50.59, to ensure the

Flood Protection Plan will be properly maintained. Therefore, the

proposed relocation of the flood protection requirements will not

increase the probability or consequences of an 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 SQN Flood Protection Plan is used to mitigate the effects of

a flooding event at SQN. This plan would not be the initiator of any

new or different kind of accident. The capability of the Flood

Protection Plan will continue to provide the same function. Changes

to the relocated requirements will be processed, in accordance with

10 CFR 50.59, to ensure the Flood Protection Plan will be properly

maintained. The proposed change does not alter the current functions

of SQN's Flood Protection Plan; therefore, this proposed change will

not create the possibility of a new or different kind of accident

from any accident previously evaluated.

B. The proposed amendment does not involve a significant

reduction in a margin of safety.

The requirements for SQN's flood protection are unchanged by the

proposed relocation of the requirements to the SQN TRM. The function

of the Flood Protection Plan and surveillance requirements to ensure

implementation of the plan remains unchanged. Any future changes to

these requirements will be evaluated, in accordance with 10 CFR

50.59, to ensure acceptability and NRC review as required.

Accordingly, the proposed change will not result in a reduction in a

margin of safety.

The NRC 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, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

NRC Project Director: Cecil O. Thomas.

[[Page 14287]]

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah

Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: February 26, 1999 (TS 99-02).

Brief description of amendments: The proposed amendments would

change the Sequoyah (SQN) Technical Specifications (TS) to provide for

consistency when exiting the action statements associated with the

Emergency Diesel Generators (D/Gs). The Tennessee Valley Authority

(TVA) inadvertently omitted revising Action Statements c, d, and e

associated with TS 3.8.1.1 in Revision 1 to TS Change 96-08, addressing

the D/G allowed outage time, submitted to the NRC staff on October 8,

1998.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), TVA 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.

This proposed revision provides for consistency and removes

contradictions within the action statements associated with TS

3.8.1.1. Additionally, the proposed revision will not result in any

change in the design, maintenance or operation of the associated

plant equipment nor will it result in deviation from the actions

presently approved by the staff for SQN's response to the associated

LCOs [Limiting Conditions for Operation]. The deletion of the

defined portion of the requirements associated with the restoration

of offsite power sources in Action Statements c and d does not

result in any change to SQN's response to the stated condition since

this requirement remains unchanged in Action Statement a.

The deletion of the requirements associated with the restoration

of 4 diesel generator (D/G) sets within 72 hours from Action

Statements c and e provides for a consistent allowed outage time of

7 days for the loss of a D/G set as previously approved by the staff

in a safety evaluation issued on December 16, 1998. Therefore, the

proposed amendment does not involve an increase in the probability

or consequences of an 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 provides for consistency and removes

contradictions within the action statements associated with TS

3.8.1.1. Additionally, the proposed revision will not result in any

change in the design, maintenance or operation of the associated

plant equipment nor will it result in deviation from the actions

presently approved by the staff for SQN's response to the associated

LCOs. Therefore, the proposed amendment does not create the

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

previously evaluated.

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed change provides for consistency and removes

contradictions within the action statements associated with TS

3.8.1.1. Additionally, the proposed revision will not result in any

change in the design, maintenance or operation of the associated

plant equipment nor will it result in deviation from the actions

presently approved by the staff for SQN's response to the associated

LCOs. Therefore, the proposed amendment does not involve a reduction

in a margin of safety.

The NRC 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, Tennessee 37402.

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.

NRC Project Director: Cecil O. Thomas.

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.

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone Nuclear

Power Station, Unit 1, New London County, Connecticut

Date of application of amendment: December 4, 1998, January 18, and

January 19, 1999.

Brief description of amendment: The proposed amendment would modify

the staffing and training requirements to allow the use of Certified

Fuel Handlers to meet plant staffing requirements.

Date of publication individual notice in Federal Register: December

29, 1998 (63 FR 71657).

Expiration date of individual notice: January 28, 1999.

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.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

[[Page 14288]]

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert Cliffs

Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

Date of application for amendment: July 20, 1998, as supplemented

December 4, 1998, and December 23, 1998.

Brief description of amendment: The amendment permits a one-time

change to the Technical Specification (TS) Bases for TS 3.8.2 for

Calvert Cliffs Nuclear Power Plant, Unit No. 2 and provides approval of

the licensee's analysis of unreviewed safety questions as described in

10 CFR 50.59. The change allows Baltimore Gas and Electric Company to

provide alternate cooling to the Unit 2 emergency diesel generators

(EDGs) during their replacement of the Unit 2 service water (SRW) heat

exchangers in the 1999 refueling outage since the normal SRW cooling

would be unavailable. The licensee proposes to provide the 2A EDG with

cooling water from the Unit 1 SRW system and to provide the 2B EDG with

cooling water from an independent external cooling system during the

replacement work.

Date of issuance: March 8, 1999.

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

during the Calvert Cliffs Unit No. 2 spring 1999 refueling outage.

Amendment No.: 205.

Facility Operating License No. DPR-69: Amendment revised the

Technical Specifications Bases.

Date of initial notice in Federal Register: August 26, 1998 (63 FR

45523) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated March 8, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

Consolidated Edison Company of New York, Docket No. 50-247, Indian

Point Nuclear Generating Unit No. 2, Westchester County, New York

Date of application for amendment: October 9, 1998.

Brief description of amendment: The amendment revised Section 6.0

to Technical Specifications to change the membership of the Nuclear

Facility Safety Committee and corrected other typographical errors.

Date of issuance: March 8, 1999.

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

30 days.

Amendment No.: 199.

Facility Operating License No. DPR-26: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 16, 1998 (63

FR 69337).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 8, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610.

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

Michigan

Date of application for amendment: April 9, 1998 (NRC-98-0071).

Brief description of amendment: The amendment revises Technical

Specification (TS) 3.7.1.2, ``Emergency Equipment Cooling Water

System,'' Action a, and TS 3.8.1.1, ``A.C. Sources--Operating,'' Action

c, to be consistent with the actions required for inoperable oxygen

monitoring instrumentation in TS 3.3.7.5, ``Accident Monitoring

Instrumentation.'' The existing ``**'' footnote to TS 3.7.1.2, Action

a, is modified and a ``*'' footnote is added to TS 3.8.1.1, Action c.

Date of issuance: March 3, 1999.

Effective date: March 3, 1999, with full implementation within 30

days.

Amendment No.: 132.

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

Technical Specifications.

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

FR 50937).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 3, 1999.

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.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-

458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

Date of amendment request: September 23, 1998.

Brief description of amendment: The amendment changes Division III

battery specific gravity acceptance criteria outlined in River Bend

Station (RBS) Technical Specifications (TS). The change is required as

a result of Division III battery system modifications scheduled to be

implemented during refueling outage RF-8, beginning April 3, 1999.

During this time, the current Division III battery will be replaced

with a new battery having a greater capacity rating. The new battery

has a nominal specific gravity of 1.215 at 77 deg.F in contrast to the

existing Division III battery supplied with a nominal specific gravity

of 1.210 at 77 deg.F. Since TS Section 3.8.6, Table 3.8.6-1 values for

specific gravity are based upon the manufacturer's nominal specific

gravity, these values were updated to reflect the changes.

Date of issuance: March 3, 1999.

Effective date: The license amendment is effective upon the date of

issuance and shall be implemented within 90 days.

Amendment No.: 103.

Facility Operating License No. NPF-47: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 18, 1998 (63

FR 64111).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 3, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear

Power Plant, Unit 1, Lake County, Ohio

Date of application for amendment: August 31, 1998.

Brief description of amendment: This amendment revised Technical

Specification Surveillance Requirement 3.6.1.3.4 to permit removal of

the inclined fuel transfer system primary containment blind flange

while primary containment integrity is required.

Date of issuance: February 24, 1999.

Effective date: February 24, 1999.

Amendment No.: 100.

Facility Operating License No. NPF-58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 21, 1998 (63 FR

56260).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 24, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

[[Page 14289]]

FirstEnergy Nuclear Operating Company, Docket No. 50-440 Perry Nuclear

Power Plant, Unit 1, Lake County, Ohio

Date of application for amendment: July 13, 1998, and as

supplemented by submittal dated November 23, 1998.

Brief description of amendment: This amendment revised Technical

Specification 3.4.4,'' Safety/Relief Valves (SRVs),'' by increasing the

present plus or minus 1% tolerance on the safety mode lift setpoint for

the safety relief valves to plus or minus 3%.

Date of issuance: March 3, 1999.

Effective date: March 3, 1999.

Amendment No.: 101.

Facility Operating License No. NPF-58: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: August 12, 1998 (63 FR

43214).

The supplemental information contained clarifying information and

did not change the initial no significant hazards consideration

determination and did not expand the scope of the original application.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 3, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Perry Public Library, 3753

Main Street, Perry, OH 44081.

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

1, DeWitt County, Illinois

Date of application for amendment: October 5, 1998.

Brief description of amendment: The amendment allows deferral of

the next scheduled local leak rate test for valve 1MC-042 until the

seventh refueling outage.

Date of issuance: March 8, 1999.

Effective date: March 8, 1999, and shall be implemented within 45

days.

Amendment No.: 121.

Facility Operating License No. NPF-62: The amendment revised the

Technical Specifications.

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

56949).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 8, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, IL 61727.

Northeast Nuclear Energy Company, Docket No. 50-245, Millstone Nuclear

Power Station, Unit 1, New London County, Connecticut

Date of application for amendment: December 4, 1998, and January 18

and 19, 1999.

Brief description of amendment: The proposed amendment would modify

the staffing and training requirements to allow the use of Certified

Fuel Handlers to meet plant staffing requirements.

Date of issuance: March 5, 1999.

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

45 days from the date of issuance.

Amendment No.: 104.

Facility Operating License No. DPR-21: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 29, 1998 (63

FR 71657).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 5, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Date of applications for amendment: August 12, 1998, as

supplemented by letter dated October 30, 1998; and application dated

September 28, 1998, as supplemented by letters dated January 7 and 20,

1999.

Brief description of amendment: The amendment allows implementation

of a revised main steamline break analysis and revised control room

habitability analyses.

Date of issuance: March 10, 1999.

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

60 days from the date of issuance.

Amendment No.: 228.

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

Technical Specifications and authorized changes to the Final Safety

Analysis Report.

Date of initial notice in Federal Register: October 7, 1998 (63 FR

53951) and December 2, 1998 (63 FR 66597).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 10, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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: December 10, 1998, as

supplemented February 19, 1999.

Brief description of amendment: The amendment allows the licensee

to implement changes to the Final Safety Analysis Report (FSAR)

regarding a revised method for ensuring boron precipitation can be

prevented (post-loss-of-coolant accident).

Date of issuance: March 10, 1999.

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

60 days from the date of issuance.

Amendment No.: 229.

Facility Operating License No. DPR-65: Amendment authorizes changes

to the Final Safety Analysis Report.

Date of initial notice in Federal Register: January 13, 1999 (64 FR

2249).

The February 19, 1999, supplemental letter provided additional

information that did not change the staff's proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 10, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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: June 10, 1998, as supplemented

October 30, 1998.

Brief description of amendment: The amendment revises the Millstone

Unit 3 licensing basis associated with post-accident mitigation

activities, vital area access travel routes, and the associated action

completion times. Northeast Nuclear Energy Company determined that the

Final Safety Analysis Report (FSAR) description of post-accident vital

area routing was out of date

[[Page 14290]]

because the radiological control area boundary fence created an access

problem on the designated routes to the hydrogen recombiner and fuel

building. The revised licensing basis will be incorporated into the

FSAR and will revise the routes to accommodate the fence location and

allow for the time to unlock gates.

Date of issuance: March 1, 1999.

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

60 days from the date of issuance.

Amendment No.: 166.

Facility Operating License No. NPF-49: Amendment authorized

revision to the FSAR.

Date of initial notice in Federal Register: July 15, 1998 (63 FR

38202).

The October 30, 1998, letter provided clarifying information that

did not change the scope of the June 10, 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 March 1, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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: December 4, 1998.

Brief description of amendment: The amendment eliminates the need

to cycle the plant and its components through a shutdown-startup cycle

by allowing the next snubber surveillance interval to be deferred until

the end of refueling outage 6 or September 10, 1999, whichever date is

earlier.

Date of issuance: March 3, 1999.

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

30 days from the date of issuance.

Amendment No.: 167.

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

Technical Specifications.

Date of initial notice in Federal Register: December 30, 1998 (63

FR 71971).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 3, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

ThreeRivers Community-Technical College, 574 New London Turnpike,

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

49 Rope Ferry Road, Waterford, Connecticut.

PP&L, Inc., Docket No. 50-388, Susquehanna Steam Electric Station, Unit

2, Luzerne County, Pennsylvania

Date of application for amendment: August 5, 1998, as supplemented

by letter dated November 23, 1998.

Brief description of amendment: This amendment would change the

allowable values for both the core spray system and the low-pressure-

coolant injection system reactor steam dome pressure-low functions.

Date of issuance: March 4, 1999.

Effective date: As of date of issuance, to be implemented within 30

days.

Amendment No.: 155.

Facility Operating License No. NPF-22: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 1, 1999 (64 FR

4904).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 4, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Osterhout Free Library,

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

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

Generating Station, Salem County, New Jersey

Date of application for amendment: August 25, 1998, as supplemented

January 27, 1999.

Brief description of amendment: This amendment revised Technical

Specification (TS) 2.1.2, ``THERMAL POWER, High Pressure and High

Flow,'' and the Bases for TS 2.1, ``Safety Limits.'' These changes were

made to implement appropriately conservative Safety Limit Minimum

Critical Power Ratio values for the Hope Creek Generating Station Cycle

9 core and fuel designs. An administrative revision has also been made

to TS 6.9.1.9 to reflect these changes for Cycle 9.

Date of issuance: March 9, 1999.

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

within 60 days after the completion of Cycle 8.

Amendment No.: 117.

Facility Operating License No. NPF-57: This amendment revised the

Technical Specifications.

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

FR 50938).

The supplemental letter provided clarifying information that did

not change the initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 9, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,

Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New

Jersey

Date of application for amendments: September 29, 1998.

Brief description of amendments: The amendments revise Technical

Specification (TS) 3/4.9.4, ``Refueling Operations--Containment

Building Penetrations,'' to allow the use of an equivalent closure

device to satisfy the closure requirements of the containment equipment

hatch during core alterations or movement of irradiated fuel in

containment. The amendment also revises TS 3/4.9.4 to allow the use of

an equivalent closure method to satisfy the closure requirements of

containment penetrations (in addition to an isolation valve, blind

flange or manual valve) during core alterations or movement of

irradiated fuel in containment.

Date of issuance: February 26, 1999.

Effective date: Effective as of its date of issuance, to be

implemented within 60 days.

Amendment Nos.: 217 and 199.

Facility Operating License Nos. DPR-70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 21, 1998 (63 FR

56258).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 26, 1999.

No significant hazards consideration comments received: No.

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

West Broadway, Salem, NJ 08079.

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

Nuclear Power Plant, Wayne County, New York

Date of application for amendment: November 24, 1998.

Brief description of amendment: This amendment revises the Ginna

Station Improved Technical Specifications

[[Page 14291]]

description of the fuel cladding material (TS 4.2.1) and updates the

list of references provided in Specification 5.6.5 for the Core

Operating Limits Report.

Date of issuance: March 3, 1999.

Effective date: As of date of issuance, to be implemented within 30

days.

Amendment No.: 73.

Facility Operating License No. DPR-18: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 30, 1998 (63

FR 71972).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 3, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Rochester Public Library, 115

South Avenue, Rochester, New York 14610.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: July 7, 1998, as supplemented by letters

dated October 15 and October 26, 1998, and February 16, 1999. The

supplements provided clarifying information and corrected

administrative errors within the scope of the amendment request and did

not change the initial no significant hazards consideration

determination.

Brief description of amendments: The amendments revised the spent

fuel pool criticality analysis and rack utilization schemes by allowing

credit for spent fuel pool soluble boron.

Date of issuance: March 3, 1999.

Effective date: This license amendment is effective as of its date

of issuance and shall be implemented within 90 days of issuance.

Amendment Nos.: Unit 1--Amendment No. 104; Unit 2--Amendment No.

91.

Facility Operating License Nos. NPF-76 and NPF-80: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 26, 1998 (63 FR

45530).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 3, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wharton County Junior College,

J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear

Plant, Unit 2, Limestone County, Alabama

Date of application for amendment: September 8, 1998 (TS-354), as

supplemented by letter dated February 22, 1999.

Brief description of amendment: Revises the Appendix A Technical

Specifications (TS) to include provisions for enabling the Oscillation

Power Range Monitor Upscale trip function in the Average Power Range

Monitor.

Date of issuance: As of date of issuance to be implemented at the

end of the Unit 2 Cycle 10 outage scheduled to begin on April 11, 1999.

Effective date: March 5, 1999.

Amendment No.: 258.

Facility Operating License No. DPR-52: Amendment revises the TS.

Date of initial notice in Federal Register: October 7, 1998 (63 FR

53958). The supplemented letter dated February 22, 1999, did not change

the original no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated March 5, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Athens Public Library, South

Street, Athens, Alabama 35611.

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

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

Virginia

Date of application for amendments: November 10, 1998.

Brief description of amendments: The amendments revise the

Technical Specifications Sections 3.4.4 and 3.4.4.a for Unit 1, and

3.4.4 and 3.4.4.a for Unit 2, providing a clarification on the

operability requirements for pressurizer heaters and the emergency

power source for the pressurizer heaters.

Date of issuance: March 1, 1999.

Effective date: March 1, 1999.

Amendment Nos.: 217 and 198.

Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised

the Technical Specifications.

Date of initial notice in Federal Register: December 2, 1998 (63 FR

66605).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 1, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

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

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

Virginia

Date of application for amendments: October 25, 1995, as

supplemented February 5, 1999. The February 5, 1999, supplemental

letter contained clarifying information only, and did not change the

initial no significant hazards consideration determination or expand

the scope of the original Federal Register Notice.

Brief description of amendments: The amendments revise the

Technical Specifications (TS) Sections 3.4.3.2, 4.4.3.2.1.b,

4.4.3.2.1.c, 4.4.3.2.2, 4.4.9.3.d, 4.4.9.3.e, 3/4.4.2, 3/4.4.3, 3/4.4.4

and 6.8.4.g for Unit 1, and 3.4.3.2, 4.4.3.2.1.c, 4.4.3.2.2, 4.4.9.3.d,

4.4.9.3.e, 3/4.4.2, 3/4.4.3, 3/4.4.4 and 6.8.4.g for Unit 2, providing

an allowed outage time of 14 days for the prezzurizer power operated

relief valve (PORV) nitrogen accumulators, as well as provide separate

action statements for the PORV depending on the reason for the PORV

inoperability.

Date of issuance: March 2, 1999.

Effective date: March 2, 1999.

Amendment Nos.: 218 and 199.

Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised

the Technical Specifications.

Date of initial notice in Federal Register: June 5, 1996 (61 FR

28620).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 2, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-

281, Surry Power Station, Units 1 and 2, Surry County, Virginia

Date of application for amendments: September 24, 1998.

Brief description of amendments: These amendments revise the

Technical Specifications to allow the reactor trip bypass breakers to

be tested immediately after being placed in service, but prior to

commencing Reactor Protection System testing or maintenance.

Date of issuance: March 12, 1999.

Effective date: March 12, 1999.

Amendment Nos.: 219 and 219.

Facility Operating License Nos. DPR-32 and DPR-37: Amendments

change the Technical Specifications.

[[Page 14292]]

Date of initial notice in Federal Register: February 10, 1999 (64

FR 6715).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 12, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments: September 28, 1998 (TSCR 208).

Brief description of amendments: These amendments clarify the

notation definition of refueling interval ``R'' in TS Table 15.4.1-1

and add a new annual (12-month) interval ``A''.

Date of issuance: March 1, 1999.

Effective date: March 1, 1999, with full implementation within 45

days.

Amendment Nos.: 186 and 191.

Facility Operating License Nos. DPR-24 and DPR-27: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 27, 1999 (64 FR

4162).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 1, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments: October 5, 1998 (TSCR 200).

Brief description of amendments: These amendments modify TS Section

15.4.1, ``Operational Safety Review,'' by removing the requirement to

check certain environmental monitors on a monthly basis.

Date of issuance: March 2, 1999.

Effective date: March 2, 1999, with full implementation within 45

days.

Amendment Nos.: 187 and 192.

Facility Operating License Nos. DPR-24 and DPR-27: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 27, 1999 (64 FR

4163).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 2, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of application for amendments: October 7, 1998 (TSCR 207).

Brief description of amendments: These amendments incorporate

changes to the Technical Specifications to ensure the 4 kV bus

undervoltage input to the reactor trip protective function is

controlled in accordance with the design and licensing basis for the

facility. An additional administrative change removes the footnote

related to the definition of Rated Power in TS 15.1.j.

Date of issuance: March 2, 1999.

Effective date: March 2, 1999, with full implementation within 45

days.

Amendment Nos.: 188 and 193.

Facility Operating License Nos. DPR-24 and DPR-27: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 30, 1998 (63

FR 71978).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated March 2, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Yankee Atomic Electric Company, Docket No. 50-29, Yankee Nuclear Power

Station, Franklin County, Massachusetts

Date of application for amendment: October 15, 1998.

Brief description of amendment: Revises the Possession Only License

by changing the submittal interval for the Radioactive Effluent Reports

from semiannual to annual.

Date of issuance: March 5, 1999.

Effective date: March 5, 1999.

Amendment No.: 151.

Possession Only License No. DPR-3: Amendment revised the Technical

Specifications.

Date of initial notice in Federal Register: November 18, 1998 (63

FR 64128). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated No significant hazards

consideration comments received: No.

Local Public Document Room location: Greenfield Community College,

1 College Drive, Greenfield, Massachusetts 01301.

Dated at Rockville, Maryland, this 17th day of March 1999.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project Management, Office of Nuclear

Reactor Regulation.

[FR Doc. 99-7032 Filed 3-23-99; 8:45 am]

BILLING CODE 7590-01-P