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 April 10, 1999, through April 23, 1999. The

last biweekly notice was published on April 21, 1999 (64 FR 19554).

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 June 4, 1999, the licensee may file a request for a hearing with

respect to issuance of the amendment to the subject facility operating

license and any person whose interest may be affected by this

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

the scope of the

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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.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam

Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of amendment request: April 12, 1999.

Description of amendment request: The proposed one-time technical

specification (TS) change, effective through September 30, 1999,

provides a Required Action and Completion Time for the Ultimate Heat

Sink (UHS) in the event that service water temperature exceeds the

current 95 deg.F surveillance limit. It involves an allowance to

continue operation for a period of 8 hours with the UHS at a

temperature greater than the temperature limits provided in TS Limiting

Condition of Operation 3.7.8, ``Ultimate Heat Sink (UHS)'' and provides

an upper UHS temperature limit beyond which plant shutdown is required.

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:

Carolina Power & Light (CP&L) Company has evaluated the proposed

Technical Specification change and has concluded that it does not

involve a significant hazards consideration. The conclusion is in

accordance with the criteria set forth in 10 CFR 50.92. The bases

for the conclusion that the proposed change does not involve a

significant hazards consideration are discussed below.

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change does not involve any physical alteration of

plant systems, structures or components. The proposed change allows

plant operation for a short period of time when the service water

temperature exceeds 95 deg.F, requires an hourly surveillance when

service water temperature exceeds 95 deg.F, provides an upper UHS

temperature limit beyond which a plant shutdown is required, and

specifies an expiration date beyond which the current requirements

are restored. If the service water temperature is restored within

the allowed time, a plant shutdown is not required. This minimizes

plant transients, which reduces the probability of a reactor trip

and the resulting challenges to mitigating systems. A service water

temperature of up to 99 deg.F does not increase the failure rate of

systems, structures or components because the systems, structures,

and components are designed for higher temperatures than at which

they operate.

The Service Water (SW) System temperature is not assumed to be

an initiating condition of any accident evaluated in the safety

analysis report. Therefore, the allowance of a limited time for

service water temperature to be in excess of 95 deg.F does not

involve an increase in the probability of an accident previously

evaluated in the safety analysis report (SAR). The SW System

supports operability of safety related systems used to mitigate the

consequences of an accident. The service water temperature is not

expected to increase significantly beyond 95 deg.F due to the

limited time allowed by the proposed change in conjunction with the

generally slow rate of temperature increase experienced from thermal

changes in Lake Robinson. The capability of components to perform

their safety related function is not affected up to a service water

temperature of 99 deg.F with the exception of the Containment Air

Recirculation Fan Coolers. The heat removal capacity of the

Containment Air Recirculation Fan Coolers is not expected to be

significantly reduced by a small increase in service water

temperature. If heat removal is not significantly reduced,

containment pressure and leakage will not be significantly

increased, and the doses from containment leakage will not be

significantly increased. Therefore, the proposed change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated in the SAR.

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

kind of accident from any accident previously evaluated?

The proposed change does not involve any physical alteration of

plant systems, structures or components. A service water temperature

of up to 99 deg.F does not introduce new failure mechanisms of

systems, structures or components not already considered in the SAR

because the systems, structures, and components are designed for

higher temperatures than at which they operate. Therefore, the

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

previously evaluated is not created.

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

of safety?

The proposed change will allow a small increase in service water

temperature above the design basis limit for the SW System and delay

by 8 hours the requirement to shutdown the plant when the service

water system design limit is exceeded. There are design margins

associated with systems, structures and components that are cooled

by the service water system that are affected. The capability of

components to perform their safety related function is not affected

up to a service water temperature [of] 99 deg.F with the exception

of the Containment Air Recirculation Fan Coolers. The Containment

Air Recirculation Fan Coolers remove heat from containment to

mitigate containment pressure and temperature following a MSLB [main

steamline break] inside containment or a Large Break LOCA [loss-of-

coolant accident] inside containment. An increase in service water

temperature in excess of the design limit due to hot weather

conditions is expected to be small due to the limited time allowed

by the proposed change in conjunction with the generally slow rate

of temperature increase experienced from thermal changes in Lake

Robinson. Therefore, the effect on the Containment Air

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Recirculation Fan Coolers' heat removal capacity and the resulting

containment pressure and temperature is expected to be small.

Therefore, there is no significant reduction in margin of safety

associated with this change.

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: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

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

Corporate Secretary, Carolina Power & Light Company, Post Office Box

1551, Raleigh, North Carolina 27602.

NRC Section Chief: Sheri R. Peterson.

Commonwealth Edison Company, Docket No. 50-254, Quad Cities Nuclear

Power Station, Units 1, Rock Island County, Illinois

Date of amendment request: March 30, 1999.

Description of amendment request: The amendment would revise the

Quad Cities Nuclear Power Station, Unit 1 Technical Specifications (TS)

by changing the Surveillance Requirements (SR) 4.6.E.2 to allow a one-

time extension of the 18-month requirement to pressure set test or

replace one half of the Main Steam Safety Valves (MSSVs) to an interval

of 24 months.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes request a one-time change to the

surveillance requirement for the MSSVs and Target Rock S/RV [Safety

Relief Valve]. The surveillance interval between MSSVs and Target

Rock S/RV testing is not a precursor assumed in any previously

analyzed accident. Therefore, the probability of a previously

evaluated accident has not been increased.

The proposed extension is consistent with the ASME Code

requirement to test 20% of the sample population every 24 months

with all of the valves in the sample group being tested every 60

months. The proposed changes are also consistent with NUREG 1433,

Revision 1, and do not adversely affect existing plant safety

margins or the reliability of the equipment assumed to operate in

the safety analysis. Operating experience and excellent materiel

condition of the MSSVs and Target Rock S/RV support the expectation

that they will continue to perform their intended function.

Therefore, the consequences of a previously evaluated accident have

not been increased.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

No new equipment is required, nor will the MSSVs and Target Rock

S/RV be operated in a different manner during the period of the

extended surveillance interval. The proposed changes are consistent

with NUREG 1433, Revision 1, requirements for safety valve

surveillance intervals as well as the ASME Code requirements for

testing safety valves. Operating experience and superior materiel

condition of the MSSVs and Target Rock S/RV support the expectation

that they will continue to perform their intended function.

Therefore, the possibility of a new or different accident has not

been increased.

Does the change involve a significant reduction in a margin of

safety?

The proposed amendment represents an extension to the current TS

SRs that would otherwise be provided generically by the ASME Code.

The proposed changes are also consistent with NUREG-1433, Revision

1, and do not adversely affect existing plant safety margins or the

reliability of the equipment assumed to operate in the safety

analysis. The proposed changes have been evaluated and found to be

acceptable for use at Quad Cities Nuclear Power Station based on

system safety analysis requirements and operational performance. The

MSSVs and Target Rock S/RV provisions continue to be adequately

maintained during plant operation. The proposed changes to the MSSVs

and Target Rock S/RV surveillance interval do not significantly

reduce existing plant safety margins since excellent materiel

condition and acceptable surveillance test results support the

expectation that no significant degradation will occur over the

extended interval.

The proposed changes are based on NRC accepted provisions at

other operating plants that are applicable at Quad Cities Nuclear

Power Station and maintain necessary levels of system or component

reliability.

The proposed amendment for Quad Cities Nuclear Power Station

will not reduce the availability of systems required to mitigate

accident conditions.

Therefore, these changes do not involve a significant reduction

in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of amendment request: March 30, 1999.

Description of amendment request: This amendment request proposes

to change the Technical Specifications (TSs) to allow an alternate

methodology for quantifying Reactor Coolant System (RCS) leakage when

the normal RCS leakage detection system is inoperable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The current Technical Specifications require a periodic

measurement of RCS leakage. The normal method for quantifying RCS

leakage is to use the DWFDS [Drywell Floor Drain Sump] and DWEDS

[Drywell Equipment Drain Sump] flow totalizers. The proposed TS

change would allow an alternate method for quantifying RCS leakage

when a flow totalizer is not available. The proposed change has no

impact on the frequency for monitoring RCS leakage and would only be

used for a maximum of 30 days while the normal leakage monitoring

system is being restored to an operable condition. The alternate

methodology for quantifying leakage has a measurement sensitivity

that is consistent with the normal method. The proposed change does

not impact any system structure or component used to mitigate the

consequences of an accident and there will be no change in the types

or significant increase in the amounts of any effluents released

offsite.

Therefore this proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed change involves no physical modifications to any

system, structure or component used to mitigate the consequences of

an accident. The operation of the DWEDS and DWFDS are not being

altered in any way that could affect their ability to function

during an accident condition.

Therefore, the proposed changes do not create the possibility of

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

Does the change involve a significant reduction in a margin of

safety?

The current TS's require a periodic measurement of RCS leakage.

The normal method for quantifying RCS leakage is to use

[[Page 24195]]

the DWFDS and DWEDS flow totalizers. The proposed technical

specifications change would allow an alternate method for

quantifying RCS leakage when a flow totalizer is inoperable. The

proposed change has no impact on the frequency for monitoring RCS

leakage and would only be used for a maximum of 30-days while the

normal leakage monitoring system is being restored to an operable

condition. The proposed alternate methodology for quantifying

leakage has a measurement sensitivity that is consistent with the

normal method.

Therefore, these changes do not involve a significant reduction

in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units and, Rock Island County, Illinois

Date of amendment request: March 30, 1999.

Description of amendment request: This amendment request proposes

to revise license conditions in each of the respective Operating

Licenses to delete those license conditions that no longer apply, make

an editorial change in the Unit 1 license, and provide clarifying

information regarding the license condition concerning equalizer valve

restrictions.

Basis for proposed no significant hazards consideration

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

its analysis of the issue of no significant hazards consideration,

which is presented below:

Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The initial conditions and methodologies used in the accident

analyses remain unchanged. The proposed changes do not change or

alter the design assumptions for the systems or components used to

mitigate the consequences of an accident. Therefore, accident

analyses results are not impacted.

The proposed changes delete various license conditions that have

been completed, make editorial changes, and provide clarifying

information. The changes are administrative. No physical or

operational changes to the facility will result from the proposed

changes.

Therefore, this proposed amendment does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes do not affect the design or operation of

any system, structure, or component in the plant. The safety

functions of the related structures, systems, or components are not

changed in any manner, nor is the reliability of any structures ,

systems, or component reduced. The changes do not affect the manner

by which the facility is operated and do not change any facility

design feature, structure, system, or component. No new or different

type of equipment will be installed.

The proposed changes delete various license conditions that have

been completed, make editorial changes, and provide clarifying

information. The changes are administrative. No physical or

operational changes to the facility will result from the proposed

changes.

Therefore, the proposed amendment does not create the

possibility of a new or different kind of accident from any

previously evaluated.

Does the change involve a significant reduction in the margin of

safety for the following reasons:

The proposed changes are administrative in nature and have no

impact on the margin of safety of any Technical Specification. There

is no impact on safety limits or limiting safety system settings.

The changes do not affect any plant safety parameters or setpoints.

The proposed changes delete various license conditions that have

been completed, make editorial changes, and provide clarifying

information. No physical or operational changes to the facility will

result from the proposed changes.

Therefore, the proposed changes do not involve a significant

reduction in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Dixon Public Library, 221

Hennepin Avenue, Dixon, Illinois 61021.

Attorney for licensee: Ms. Pamela B. Stroebel, Senior Vice

President and General Counsel, Commonwealth Edison Company, P.O. Box

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

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

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

Date of amendment request: March 25, 1999.

Description of amendment request: The amendments would revise the

Technical Specifications (TS) to redefine the ``trip setpoint'' in a

number of locations as the ``nominal trip setpoint.'' The current

definition results in upper-or lower-bound numerical values not to be

exceeded for setpoints. This proposed new definition would permit the

setpoints to be set within a tolerance range around the number

specified in various tables. The TS locations affected are: Table

3.3.1-1, ``Reactor Trip System Instrumentation;'' Table 3.3.2-1,

``Engineered Safety Feature Actuation Instrumentation;'' Surveillance

Requirement 3.3.5.2; Table 3.3.6-1, ``Containment Purge and Exhaust

Isolation Instrumentation;'' and Limiting Condition of Operation (LCO)

3.4.12. Sections of the associated TS Bases document would also be

revised to reflect the TS changes.

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?

No. The proposed changes are consistent with the current

licensing basis for Catawba Nuclear Station, the setpoint

methodology used to develop the Trip Setpoints, the Catawba Safety

Analyses, and current station calibration procedures and practices.

The Reactor Trip System and Engineered Safety Features Actuation

System are not accident initiating systems; they are accident

mitigating systems. Therefore, these proposed changes will have no

impact on any accident probabilities. Accident consequences will not

be affected, as no changes are being made to the plant which will

involve a reduction in reliability of these systems. Consequently,

any previous evaluations associated with accidents will not be

affected by these changes.

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

kind of accident from any accident previously evaluated?

No. The proposed changes are consistent with the current

licensing basis for Catawba Nuclear Station, the setpoint

methodology used to develop the Trip Setpoints, the Catawba Safety

Analyses, and current station calibration procedures and practices.

No changes are being made to actual plant hardware which will result

in any new accident causal mechanisms. Also, no changes are being

made to the way in which the plant is being operated. Therefore, no

[[Page 24196]]

new accident causal mechanisms will be generated. Consequently,

plant accident analyses will not be affected by these changes.

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

of safety?

No. The proposed changes are consistent with the current

licensing basis for Catawba Nuclear Station, the setpoint

methodology used to develop the Trip Setpoints, the Catawba Safety

Analyses, and current station calibration procedures and practices.

Margin of safety is related to the confidence in the ability of the

fission product barriers to perform their design functions during

and following accident conditions. These barriers include the fuel

cladding, the reactor coolant system, and the containment system.

The performance of these barriers will not be degraded by the

proposed changes. Consequently, plant safety analyses will not be

affected by these changes.

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

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

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

amendment request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina.

Attorney for licensee: Ms. Lisa F. Vaughn, Legal Department

(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte,

North Carolina.

NRC Section Chief: Richard L. Emch, Jr.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of amendment request: April 9, 1999.

Description of amendment request: The proposed amendment would

modify the Technical Specifications (TSs) to add Limiting Condition for

Operation (LCO) 3.0.6 and its associated bases. This change would allow

equipment that has been removed from service or declared inoperable in

compliance with the TS Action statement to be returned to service under

administrative controls solely to perform testing required to

demonstrate its operability or the operability of other equipment. The

proposed change is consistent with TS 3.0.5 as discussed in NUREG-1432,

Revision 1, ``Standard Technical Specifications for Combustion

Engineering Plants.'' TS 3.0.2 would also be modified to reflect that

TS 3.0.6 is an exception to TS 3.0.2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

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

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

presented below:

1. The proposed change does not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed change would allow an orderly return to service of

inoperable equipment. This change does not alter the functional

characteristics of any plant component and does not allow any new modes

of operation of any component. The accident mitigation features of the

plant are not affected by the proposed amendment request. Therefore,

this proposed amendment would not result in a significant change in the

types or significant increase in the amounts of any effluents that may

be released off site. No modifications to the plant have been proposed

due to this amendment request. The proposed change would permit

equipment removed from service to comply with required actions to be

returned to service under administrative controls to verify the

operability of the equipment being returned to service or of other

related equipment. Although returning inoperable equipment to service

for testing may temporarily compromise single failure criteria,

administrative controls will ensure the time involved will be limited

to only that required to demonstrate component or system operability.

This LCO provides an acceptable method of restoring equipment to

service for the sole purpose of demonstrating its operability or the

operability of other related equipment. Therefore, this change does not

involve a significant increase in the probability or consequences of

any accident previously evaluated.

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

different kind of accident from any previously evaluated.

No modifications were made to the plant due to this amendment

request. The proposed change does not alter the functional

characteristics of any plant component and does not allow any new modes

of operation for any component. This proposed amendment would

facilitate the testing of equipment in its design configuration to

demonstrate operability. The use of TS 3.0.6 would be limited to the

time absolutely necessary to perform the test. Therefore, this change

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

from any previously evaluated.

3. The proposed change does not involve a significant reduction in

a margin of safety.

The addition of TS 3.0.6 is considered necessary to establish an

allowance that is not formally recognized in the current TSs. Without

this allowance, situations can arise in which certain components could

not be restored to operable status without requiring a plant shutdown.

It is not the intent that the TSs preclude the return to service of a

component to confirm its operability. This allowance is deemed to

represent a more stable, safe operation than requiring a plant shutdown

to complete the restoration and confirmatory testing. The time period

during which the equipment is returned to service in conflict with the

requirements of the TS Action statement is limited to the time

absolutely necessary to perform the indicated surveillance requirement.

TS 3.0.6 does not provide time to perform any other preventive or

corrective maintenance. The period of time during which the equipment

is returned to service will be limited by administrative controls and

is considered very small. Therefore, the probability of an accident

during that time period is also very small and is considered to be

insignificant. Thus, it can be concluded that the proposed change does

not affect the current 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: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801.

Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and

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

NRC Section Chief: Robert A. Gramm.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse

Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: March 9, 1999.

Description of amendment request: The proposed change would modify

the Technical Specifications to increase the inservice inspection

interval, and reduce the scope of volumetric and surface examinations

for the reactor coolant pump flywheels.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis of the issue of no significant hazards

consideration, which is presented below:

The Nuclear Regulatory Commission has provided standards in 10

CFR 50.92(c) for

[[Page 24197]]

determining whether a significant hazard exists due to a proposed

amendment to an Operating License for a facility. A proposed

amendment involves no significant hazards consideration if operation

of the facility in accordance with the proposed changes would: (1)

Not involve a significant increase in the probability or

consequences of an accident previously evaluated; (2) Not create the

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

previously evaluated; or (3) Not involve a significant reduction in

a margin of safety. The Davis-Besse Nuclear Power Station has

reviewed the proposed changes and determined that a significant

hazards consideration does not exist because operation of the Davis-

Besse Nuclear Power Station, (DBNPS) Unit No. 1, in accordance with

these changes would:

1a. Not involve a significant increase in the probability of an

accident previously evaluated because no accident initiators,

conditions, or assumptions are affected by the proposed changes to

Technical Specification Surveillance Requirement 4.4.10.1.a in the

frequency and scope of volumetric and surface examinations for the

Reactor Coolant Pump (RCP) motor flywheels.

1b. Not involve a significant increase in the consequences of an

accident previously evaluated because changes in the frequency and

scope of volumetric and surface examinations for the RCP motor

flywheels will not affect any previously evaluated accidents.

Accidents associated with failure of the flywheel were not evaluated

in the DBNPS Updated Safety Analysis Report (USAR). The design,

fabrication, and testing of flywheels in accordance with the

guidance found in NRC Regulatory Guide 1.14, ``Reactor Coolant Pump

Flywheel Integrity,'' Revision 1, August 1975, minimizes the

potential for flywheel failure. The proposed changes have been

demonstrated to maintain conservative testing requirements for the

flywheels.

2. Not create the possibility of a new or different kind of

accident from any previously evaluated because changes in the

frequency and scope of volumetric and surface examinations for the

RCP motor flywheels will not affect the reliability of RCP motor

flywheels. No new failure mode is introduced since the proposed

changes do not involve a modification or change in operation of any

plant systems, structures, or components.

3. Not involve a significant reduction in the margin of safety.

As shown in Westinghouse Topical Report WCAP-14535A, ``Topical

Report on Reactor Coolant Pump Flywheel Inspection Elimination,''

November 1996, RCP motor flywheels have been inspected for twenty

years without any service induced flaws being identified.

Additionally, the analyses demonstrated that the flywheels are

manufactured from high quality steel, have a high fracture

toughness, and have a very high flaw tolerance. The topical report

indicates that the flywheels could be operated for forty years

without inspection, and there would be no significant increase in

the probability of failure of the flywheels. However, inspections

are proposed to continue at a frequency of once every ten years as a

conservative measure. Thus, the margin of safety is not reduced

significantly by the proposed change in inspection frequency.

Based on the above, the Davis-Besse Nuclear Power Station has

determined that the License Amendment Request does not involve a

significant hazards consideration. As this License Amendment Request

concerns a proposed change to the Technical Specifications that must

be reviewed by the Nuclear Regulatory Commission, this License

Amendment Request does not constitute an unreviewed safety question.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, OH 43606.

Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,

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

NRC Section Chief: Anthony J. Mendiola.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: March 31, 1999.

Description of amendment request: The proposed change would modify

Cooper Nuclear Station's technical specification administrative

controls for unit staff qualifications for the shift supervisor, senior

operator, licensed operator, shift technical advisor, and radiological

manager.

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 does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. The proposed change provides enhancement to the current

requirements and clarifies the qualifications and training

requirements for the shift supervisor, senior operator, licensed

operator, shift technical advisor, and Radiological Manager. This

provides additional assurance that these personnel are properly

trained and qualified for their positions; therefore, the proposed

change does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change will not create the possibility of a new or

different kind of accident than evaluated in the Updated Safety

Analysis Report (USAR). The proposed change provides enhancement to

the current requirements and clarifies the qualifications and

training requirements for the shift supervisor, senior operator,

licensed operator, shift technical advisor, and Radiological

Manager. The revised administrative controls for unit staff

qualifications are an enhancement to the current requirements;

therefore, the proposed change does not create the possibility of a

new or different kind of accident.

The proposed change will not create a significant reduction in

the margin of safety. The proposed change provides enhancement to

the current requirements and clarifies the qualifications and

training requirements for the shift supervisor, senior operator,

licensed operator, shift technical advisor, and Radiological

Manager. This provides additional assurance that these personnel are

properly trained and qualified for their positions; therefore, the

proposed change will not create 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: Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE 68305.

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus, NE 68602-0499.

NRC Section Chief: Robert A. Gramm.

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

Nuclear Station, Unit 2, Oswego County, New York

Date of amendment request: March 31, 1999.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Table 3.6.1.2-1, ``Allowable Leak

Rates through Valves in Potential Bypass Leakage Paths,'' by adding two

relief valves, with associated leak rate criteria, to be installed on

the drywell equipment drain line and drywell floor drain line during

the refueling outage in the spring of 2000. Specifically:

(i) For the drywell equipment drain line, the reference to the

inboard isolation valve (2DER*MOV119) would be replaced with a

reference to the isolation valve and its associated relief valve

(2DER*MOV119 and 2DER*RV344);

(ii) For the drywell floor drain line, the reference to the inboard

isolation valve (2DFR*MOV121) would be

[[Page 24198]]

replaced with a reference to the isolation valve and its associated

relief valve (2DFR*MOV121 and 2DFR*RV228); and

(iii) A footnote for both above changes would be added to state,

``For valves 2DER*MOV 119 and 2DER*RV344, and likewise for valves

2DFR*MOV121 and 2DFR*RV228, this limit shall be the combined allowable

leak rate and not the per valve allowable leak rate.''

The two relief valves would be installed to protect the drain line

penetrations against overpressure, consistent with Generic Letter 96-

06, ``Assurance of Equipment Operability and Containment Integrity

During Design-Basis Accident Conditions.'' The allowable leak rates

currently specified in TS Table 3.6.1.2-1 for the drywell equipment and

drywell floor drain line penetrations will not be increased as a result

of the hardware modifications or proposed TS amendment.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The operation of Nine Mile Point Unit 2 [NMP2], in accordance

with the proposed amendment, will not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed amendment will add one relief valve on the drywell

equipment drain line (penetration 2DER*Z40) and one relief valve on

the drywell floor drain line (penetration 2DFR*Z39). These valves

will be installed on piping between the inboard containment

isolation valve and the primary containment wall. These drain lines

represent potential bypass leakage paths from the primary

containment to the environment and are subject to maximum allowable

isolation valve leak rates, as specified in Table 3.6.1.2-1 of the

Technical Specifications (TS). The purpose of adding relief valves

is to protect the piping between the inboard and outboard isolation

valves against thermally induced overpressure under postulated

accident conditions when both isolation valves close, and the fluid

trapped between them may heat up and expand. The new relief valves

and piping will not cause any existing plant design, operating, or

testing limits to be exceeded. The relief valve installations will

meet standards and specifications currently applicable to the

penetrations being modified. The relief valve configuration, set

pressure, and testing meet applicable NRC guidance. No different

precursors or new accident initiators are introduced as the result

of the proposed modification. Therefore, this proposed amendment

does not involve a significant increase in the probability of an

accident previously evaluated.

The existing requirements relating to allowable bypass leakage

for the two penetrations affected by this modification, will not be

changed. No new bypass leakage paths to the environment will be

created and no new failure modes will be introduced. Should the

relief valves open and fail to close, the effectiveness of the

containment and other fission product barriers will not be

compromised. As a result, accident dose rates will remain unchanged

and within the limits of 10 CFR 50, Appendix A, General Design

Criterion 19, and 10 CFR 100. None of the accident assumptions

described in Section 6.2, titled ``Containment Systems'' and Chapter

15, titled ``Accident Analysis,'' of the NMP2 Updated Safety

Analysis Report (USAR) is adversely affected by the proposed

modifications. Therefore, this proposed amendment does not involve a

significant increase in the consequences of an accident previously

evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

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

different kind of accident from any accident previously evaluated.

The isolation valves associated with penetrations 2DER*Z40

(drywell equipment drain line) and 2DFR*Z39 (drywell floor drain

line) perform an accident mitigation function by isolating the

containment during and after certain postulated accidents. The

addition of relief valves between the inboard and outboard isolation

valves will enhance the capability of the existing isolation valves

to perform their function without the risk of failure due to piping

overpressurization. Consistent with the guidance in Generic Letter

96-06, the consequences of a stuck-open relief valve malfunction

have been evaluated and are acceptable. Should the relief valve fail

to close after opening, the existing outboard isolation valve will

perform its function to isolate the containment. Therefore,

operation of NMP2 in accordance with this proposed amendment will

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

from any accident previously evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

The proposed installation of the relief valves will not

adversely affect primary containment integrity, the maximum

allowable leak rates for the affected penetrations, any other

fission product barriers, or any plant safety/operational limits.

The relief valves will assure that the associated isolation valves

do not fail as the result of piping overpressure during and after

postulated accidents, which will preserve the radiological margin of

safety. Therefore, operation of NMP2 in accordance with the proposed

amendment will not involve a significant reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

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

NRC Section Chief: S. Singh Bajwa

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-423,

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

Connecticut

Date of amendment request: March 2, 1999.

Description of amendment request: The proposed amendment would

require two service water (SW) pumps and their associated strainers to

be operable to declare a service water system (SWS) loop operable. The

proposed amendment would also (1) modify the existing action statement

to take into account one or more service water pump(s) or strainers

being inoperable and (2) make changes to the appropriate Bases section.

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:

NNECO has reviewed the proposed revision in accordance with 10

CFR50.92 and has concluded that the revision does not involve any

Significant Hazards Considerations (SHC). The basis for this

conclusion is that the three criteria of 10CFR50.92(c) are not

satisfied. The proposed Technical Specification revision does not

involve an SHC because the revision would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed TS [Technical Specification] change adds an

additional AOT [allowed outage time] for one of four of the service

water pumps/strainers in the SWS. The capabilities of the SWS were

evaluated in order to ensure that a significant increase in the

probability or consequences of the following previously evaluated

accidents, LOP [loss of power], LOCA [loss-of-coolant accident] with

concurrent LOP and secondary side piping break inside containment,

are precluded by SWS mitigative functions. As the above DBA's

[design basis accidents] are not caused by the failure of the SWS to

operate, the SWS can not affect the probability of these accidents

to occur.

Since both pumps/strainers in each loop are covered by the

ACTION statement in the TS when inoperable (due to failure or

maintenance), and the proposed ACTION statement for two inoperable

service water pumps in a single loop is consistent with the

[[Page 24199]]

current ACTION statement, there is no impact on the capability to

maintain core decay heat removal following a DBA. Further, the

revised TS will improve availability of the SWS. The LCO [limiting

condition for operation] and ACTION statements help ensure that the

SWS, including pumps/strainers, are kept in a condition which allows

it to perform all its design functions including providing core

decay heat removal and the SFP [spent fuel pool] cooling. As such,

there is no affect on the consequences of previously evaluated

accidents.

Thus, it is concluded that the proposed revision does 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 accident previously evaluated.

The SWS is used to remove heat from the reactor plant auxiliary

systems and other systems. Only one of four pumps is required to be

operating during normal plant conditions. In addition, only one 100%

capacity pump is required to provide the necessary flow to mitigate

the consequences of a DBA. This change continues to require two

pumps/strainers per loop to be operable and imposes strict controls

on the AOT for the SWS pumps/strainers via the imposition of the LCO

controls on the SWS. This assures that four service water pumps/

strainers will always be available or the plant will be in an ACTION

STATEMENT. The SWS is used to mitigate the consequences of an

accident and will not cause an accident.

Thus, this proposed revision does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

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

This change will have no impact on the performance of any safety

related system covered by the TS. This change explicitly defines the

number of pumps/strainers required for the SWS to be considered

OPERABLE and the ACTION required which specifies the AOT for

inoperable components. The required flow rate for accident

mitigation continues to be available to all ECCS [emergency core

cooling system] components and their support systems. As such, this

change does not increase the peak clad temperature for a DBA-LOCA.

The proposed Technical Specification change adds an additional

AOT for one of four of the service water pumps/strainers in the SWS.

Two service water pumps/strainers are required to perform the design

function of the SWS; one pump to mitigate the DBA and the other to

reduce the potential of the SFP boiling which could occur if a

service water pump is unavailable for SFP cooling after a design

basis LOCA.

The existing TS Bases states that ``The OPERABILITY of the

Service Water System ensures that sufficient cooling capacity is

available for continued operation of safety-related equipment during

normal and accident conditions. The redundant cooling capacity of

this system, assuming a single failure, is consistent with the

assumptions used in the safety analyses.''

Since this change continues to control the availability of the

SW pumps by placing the system in an ACTION statement with one loop

out of service, then the change will continue to comply with the

existing BASES requirements. Thus it is concluded that the proposed

revision does not involve a significant reduction in the margin of

safety.

In conclusion, based on the information provided, it is

determined that the proposed revision does not involve a SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Section Chief: James W. Clifford.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic City Electric Company, Dockets

Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2

and 3, York County, Pennsylvania

Date of application for amendments: December 24, 1998.

Description of amendment request: Revises the setpoints and limits

of allowable values for loss of power (LOP) instrumentation for 4kV

emergency busses.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The LOP instrumentation provides safety-related electrical

equipment protection. No new equipment is added to the plant as a

result of the proposed changes. Separation of the 4kV emergency

buses from the grid is the only potential transient that previously

existed based on operation of these relays. Based on the revised

Voltage Regulation Study, which incorporates the effects of system

improvements and additional conservatisms, there is no significant

increase in the probability of this separation. The relay time delay

settings are such that the relays will detect and respond to an

actual sustained degradation of voltage, but will not actuate in

response to normal operational voltage fluctuations. No accident

initiators will be impacted by the proposed setpoint changes. All

safety systems will be able to perform their safety functions.

Accident mitigation is achieved by these relays by ensuring adequate

voltage is maintained throughout the Class 1E electrical

distribution system.

The existing allowable values and the proposed allowable values

for Functions 2, 3, 4, and 5 have been analyzed and both values are

acceptable for operation. During implementation of modification 96-

01511 (changing of the relay setpoints), the 4kV buses could be in

one of the three configurations: (a) Both sources have relays set at

the existing setpoints, (b) one set of source relays with the

existing old setpoints and the other set with the proposed revised

setpoints, or (c) both sources have relays set at the proposed

revised setpoints. Each of these configurations is acceptable

because the existing and proposed values satisfy the design limits

established within the setpoint calculation and the Voltage

Regulation Study.

For Function[s] 4 and 5, the present TS has separate entries in

Table 3.3.8.1-1, for the internal and external time delay. This

proposed change will combine these internal and external time delays

for simplicity. The aggregate time delay is the important parameter

and it is the only time delay that is analyzed. The internal time

delay minimizes the relay contact wear and reduces the number of

external time delay relay actuations due to transient voltage dips.

The internal time delay provides no other output functions.

Therefore, there will be no impact on the Class 1E power

distribution system to perform its intended design function.

Therefore, the proposed changes described above, or operation

while modification 96-01511 is being implemented, does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

2. The proposed changes do not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed LOP instrumentation setpoint changes will not

result in any new accidents or operational transients. Separation of

the 4kV emergency buses from the grid is the only potential

transient that previously existed based on operation of these

relays. Based on the revised Voltage Regulation Study, which

incorporates the effects of system improvements and additional

conservatisms, there is no significant increase in the probability

of this separation, and the proposed setpoint changes would not

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

any previously evaluated. The relay time delay settings are such

that the relays will detect and respond to an actual sustained

degradation of voltage, but will not actuate in response to normal

operational voltage fluctuations. The proposed setpoint changes for

these relays and the proposed combining

[[Page 24200]]

of the internal and external time delays will not become initiators

of different types of accidents or transients. Additionally, since

the existing and proposed allowable values for the LOP

instrumentation functions are within the band established by the

Voltage Regulation Study, both values are acceptable for operation

during the implementation of modification 96-01511. Therefore, the

possibility of a new or different kind of accident than previously

evaluated is not created.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

All LOP instrumentation functions will continue to be carried

out. The proposed setpoint and allowable value changes have been

evaluated within the Voltage Regulation Study and the Plant

Electrical Load Study. The relay setpoints have been established

using IISCP setpoint methodology. The setpoint determination

accounts for relay accuracy, potential transformer accuracy,

measurement and test equipment accuracy, and margin above the design

limit established within the Voltage Regulation Study. The proposed

setpoint changes for these relays and the proposed combining of the

internal and external time delays will not involve a significant

reduction in a margin of safety. Additionally, since the existing

and proposed allowable values for the LOP instrumentation functions

are within the band established by the Voltage Regulation Study,

both values are acceptable for operation during the implementation

of modification 96-01511. Therefore, having both values during the

implementation of modification 96-01511 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: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

PA 17105.

Attorney for Licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and

General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,

PA 19101.

NRC Project Director: Elinor G. Adensam

PECO Energy Company, Public Service Electric and Gas Company, Delmarva

Power and Light Company, and Atlantic City Electric Company, Dockets

Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Units Nos. 2

and 3, York County, Pennsylvania

Date of application for amendments: February 12, 1999.

Description of amendment request: Administrative changes to correct

typographic errors in Technical Specifications (TS) introduced in

previous amendments.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed changes correct typographical errors and are

administrative only and do not impact the operation of the facility.

In each case, the action of the intended TS requirements were

satisfactorily completed when the change was implemented. These

corrections are administrative only and have no effect on any

previously evaluated accident scenario. The changes will not alter

the operation of equipment assumed to be available for the

mitigation of accidents or transients, nor will they alter the

operation of equipment important to safety previously evaluated.

Therefore, the changes do not involve a significant increase in

the probability or consequences of an accident previously evaluated.

2. The proposed changes do not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed changes correct typographical errors and are

administrative only and will not involve any physical changes to the

plant SSCs [systems, structures, or components]. In each case, the

action of the intended TS requirements were satisfactorily completed

when the change was implemented. These corrections are

administrative only and have no effect on any previously evaluated

accident scenario. The proposed changes do not allow operation in

any mode that is not already evaluated. The changes will not alter

the operation of equipment important to safety previously evaluated.

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

different kind of accident from any accident previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed changes correct typographical errors and are

administrative only and will not affect the manner in which the

facility is operated, or change equipment or features which affect

the operational characteristics of the facility. The proposed

changes have no impact on any safety analysis assumptions or margins

of safety.

Therefore, these 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: Government Publications

Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

PA 17105.

Attorney for Licensee: J.W. Durham, Sr., Esquire, Sr. V.P. and

General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,

PA 19101.

NRC Section Chief: James W. Clifford.

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

Oglethorpe Power Corporation, Municipal Electric Authority of Georgia,

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

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

Date of amendment request: January 21, 1999.

Description of amendment request: The proposed amendments would

change Technical Specification Tables 3.3.6.1-1 and 3.3.6.2-1 by

increasing the Allowable Values for the high radiation trip for the

exhaust monitors for the reactor building and the refueling. The

January 21, 1999, amendment request supercedes the July 22, 1998,

amendment request which was noticed in the Federal Register on August

26, 1998 (63 FR 45529).

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). Do the proposed changes involve a significant increase in

the probability or consequences of an accident previously evaluated?

The Unit 1 and Unit 2 reactor building and refueling floor

ventilation exhaust radiation monitors perform no function in

preventing, or decreasing the probability of, a previously evaluated

accident. The monitors are designed to monitor ventilation exhaust

for indications of a release of radioactive material resulting from

a design basis accident and initiate appropriate protective actions.

Because the proposed changes affect only the ventilation exhaust

radiation monitors, the probability of an accident previously

evaluated remains the same.

The function of the reactor building and the refueling floor

ventilation exhaust radiation monitors, in combination with other

accident mitigation systems, is to limit fission product release

during and following postulated design basis accidents. The proposed

new Allowable Values for the high radiation trip will continue to

ensure the offsite doses resulting from a design basis accident do

not exceed the NRC-approved

[[Page 24201]]

licensing basis. Therefore, the proposed changes do not involve a

significant increase in the consequences of an accident previously

evaluated.

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

different kind of accident from any accident previously evaluated?

The proposed changes increase the radiation level at which the

ventilation exhaust monitors actuate; however, the manner in which

their actuation logic functions and the systems that isolate or

actuate as a result are unaffected by the proposed changes.

Furthermore, the ventilation exhaust monitors will continue to

perform their design function of limiting offsite doses to NRC-

approved licensing limits at the higher Allowable Values. Therefore,

the proposed 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 Bases for Unit 1 and Unit 2 Technical Specifications Tables

3.3.6.1-1 and 3.3.6.2-1 state that the Allowable Values for the

reactor building and refueling floor ventilation exhaust radiation

monitors ``are chosen to ensure radioactive releases do not exceed

offsite dose limits.'' The proposed Allowable Values ensure the

radiation monitors actuate at a radiation level sufficient to ensure

offsite doses are within the NRC-approved licensing basis. The

proposed Allowable Values comply with the margin of safety defined

in the Technical Specifications Bases for the ventilation exhaust

radiation monitors; therefore, the proposed changes do not reduce 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: Appling County Public Library,

301 City Hall Drive, Baxley, Georgia.

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

Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC.

NRC Section Chief: Richard L. Emch, Jr.

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: March 30, 1999.

Description of amendment request: The licensee has proposed to

relocate Technical Specification 3/4.3.3.4, ``Meteorological

Instrumentation,'' and its associated Bases to the Technical

Requirements Manual (TRM). Because the TRM is incorporated within the

South Texas Project updated final safety analysis report (UFSAR) for

the units, changes to the requirements on the meteorological

instrumentation that would be relocated to the TRM would be controlled

in accordance with 10 CFR 50.59.

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 affected system and components [i.e., meteorological

monitoring instrumentation] are not assumed as initiators of

analyzed events, and are not assumed to mitigate accident or

transient events. The requirements and surveillances for [this

affected system] and components will be relocated from the Technical

Specifications to the Technical Requirements Manual, which is

incorporated in the South Texas Project UFSAR and will be maintained

pursuant to 10 CFR 50.59. In addition, the Meteorological Monitoring

System components are addressed in existing surveillance procedures

which are also controlled by 10 CFR 50.59 and subject to the change

control provisions imposed by plant administrative procedures, which

endorse applicable regulations and standards. The associated changes

to the Technical Specification Index are administrative. Therefore,

this change does not involve a significant increase in the

probability or 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.

This change does not involve a physical alteration of the plant

(no new or different type of equipment will be installed) or make

changes in the methods governing normal plant operation. This change

will not impose different requirements, and adequate control of

information will be maintained. Furthermore, this change will not

alter assumptions stated in the safety analysis or licensing basis.

The associated changes to the Technical Specification Index are

administrative. Therefore, this change does not create the

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

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

This change will not reduce a margin of safety because the

change has no impact on any safety analysis assumptions. In

addition, the relocated requirements and surveillances for the

affected structures, systems, and components remain the same as the

existing Technical Specifications. Because any future changes to

these requirements or the surveillance procedures will be evaluated

per the requirements of 10 CFR 50.59; there is no [significant]

reduction in a margin of safety. The associated changes to the

Technical Specification Index are administrative and have no

potential effect on the margin of safety. Therefore, this 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 standards of 10 CFR 50.92(c) are

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas

77488.

Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, NW., Washington, DC 20036-5869.

NRC Section Chief: Robert A. Gramm.

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

Nuclear Plant (SQN), Units 1 and 2, Hamilton County, Tennessee

Date of application for amendments: March 2, 1999 (TS 98-05).

Brief description of amendments: The proposed amendments would

change the SQN Operating Licenses DPR-77 (Unit 1) and DPR-79 (Unit 2)

by eliminating a requirement to have an Independent Safety Engineering

Group (ISEG), conditions imposed by NUREG-0737. Because of evolution

through numerous reorganizations and reassignments, these license

conditions are no longer necessary and the Tennessee Valley Authority

(TVA, the licensee) proposes deleting them.

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), 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 possibility of occurrence or the consequences for an

accident or malfunction of equipment is not increased. The ISEG

function is one of ``oversight'' only.

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

new or different kind of accident from any accident previously

evaluated.

A possibility for an accident or malfunction of a different type

than any evaluated previously in SQN's Final Safety Analysis Report

is not created by the proposed elimination of the ISEG; nor is the

possibility for an accident or malfunction of a different type. The

ISEG function is one of ``oversight'' only.

[[Page 24202]]

C. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed amendment will not involve a significant reduction

in the margin of safety. The ISEG function is one of ``oversight''

only.

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 Section Chief: Sheri R. Peterson.

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

Electric Station, Units 1 and 2, Somervell County, Texas

Date of amendment request: February 12, 1999 (TXX-99022).

Brief description of amendments: The proposed changes would modify

the steam generator tube inspection requirements and acceptance

criteria to implement the 1.0-volt repair criteria for steam generator

tubes affected by outer diameter stress corrosion cracking (ODSCC)

according to Nuclear Regulatory Commission (NRC) Generic Letter 95-05

(``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes

Affected by Outside Diameter Stress Corrosion Cracking'') at Comanche

Peak Unit 1. Also proposed is the use of a voltage-dependent

probability of detection; the methodology was originally submitted to

the NRC by the Nuclear Energy Institute in 1996.

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 Comanche Peak Unit 1 in accordance with the

proposed license amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Tube burst criteria are inherently satisfied during normal

operating conditions due to the proximity of the tube support plate

[TSP]. Test data indicates that tube burst cannot occur within the

TSP, even for tubes which have 100% through-wall electric discharge

machining notches, 0.75 inch long, provided that the TSP is adjacent

to the notched area. Since tube to tube support plate proximity

precludes tube burst during normal operating conditions, use of the

criteria must retain tube integrity characteristics which maintain a

margin of safety of 1.43 times the bounding faulted condition (Steam

Line Break) pressure differential. As previously stated, the RG

[Regulatory Guide] 1.121 criterion requiring maintenance of a safety

factor of 1.43 times the Steam Line Break pressure differential on

tube burst is satisfied by \3/4\'' diameter tubing with bobbin coil

indications with signal amplitudes less than 4.7 volts, regardless

of the indicated depth measurement. At the FDB [flow distribution

baffle], a safety factor of 3 against the normal operating condition

at power is applied. Here a voltage of 3.34 volts satisfies the

burst capability recommendation.

The upper voltage repair limit (VURL) will be

determined prior to each outage using the most recently approved NRC

database to determine the tube structural limit (VSL).

The structural limit is reduced by allowances for nondestructive

examination (NDE) uncertainty (VNDE) and growth

(VGr) to establish VURL. As an example, the

NDE uncertainty component of 20% and a voltage growth allowance of

30% per full power year can be utilized to establish a

VURL of 3.13 volts for TSP indications, 2.22 volts for

the FDB indications. The 20% NDE uncertainty represents a

squareroot-sum-of-the-squares (SRSS) combination of probe wear

uncertainty and analyst variability.

The flaw growth allowance should be an average growth rate or

30% per effective full power year, whichever is larger. The 30%

growth allowance used to determine VURL is conservative

for the current conditions at Comanche Peak Unit 1. The average

growth of the bobbin indication voltages observed at the last

inspection is determined to be 0.14 volts, or 24.6% voltage growth.

This value is a conservative representation of the growth trends at

Comanche Peak Unit 1 as not all steam generators were inspected at

end of cycle 3 and end of cycle 4, and the largest reported voltage

growths represent more than one cycle of actual plant operation. The

most current NRC approved database, contained in EPRI [Electric

Power Research Institute] NP-7480-L, Addendum 1, was used to

establish the VURL values for the FDB and TSP

intersections. Once approved by the NRC, the industry protocol for

updating the database will be followed by TU Electric, ensuring that

the most current database is utilized for all future applications of

the criteria.

Also, assuming the criteria was applied at the last inspection

at Comanche Peak Unit 1, using conservative growth projections as

described in Reference 2 [of the February 12, 1999, application],

the conditional burst probability at end of cycle 6 is determined to

be 1.7 x 10-4, which is well within the GL 95-05

reporting limit of 1 x 10-2.

Relative to the expected leakage during accident condition

loadings, it has been previously established that a postulated main

Steam Line Break outside of containment but upstream of the MSIV

[main steam isolation valve] represents the most limiting

radiological condition relative to the plugging criteria. In support

of implementation of the revised plugging limit, it will be

determined whether the distribution of cracking indications at the

tube support plate intersections during future cycles are projected

to be such that primary to secondary leakage would result in site

boundary doses within 10CFR100 guidelines and control room doses

within the GDC [General Design Criterion]-19 limit. A separate

calculation has determined this allowable Steam Line Break leakage

limit to be 27.79 gpm in the faulted loop assuming a RCS [reactor

coolant system] dose equivalent I-131 concentration of 1.0 microCi/

gm. The establishment of the 27.79 gpm leak rate value is controlled

by the 0 to 2 hour offsite dose at the site boundary for the

accident initiated iodine spike case, not the control room dose. For

this case, the site boundary thyroid dose approaches, but is bounded

by, the 30 Rem limit recommended in NUREG-0800 [``Standard Review

Plan''].

The methods for calculating the radiological dose consequences

are also revised for this application. Rather than basing the

calculated thyroid dose consequences on conversion factors from TID-

14844, [``Calculation of Distance Factors for Power and Test Reactor

Sites''] factors obtained from ICRP-30 [International Commission on

Radiation Protection Publication 30] are used. The use of ICRP-30

dose conversion factors in this application has been previously

accepted by the NRC. Although the use of ICRP-30, relative to the

TID-14844, results in lower calculated thyroid doses for this

application, the NRC has previously determined that the ICRP-30

factors retain adequate conservatism.

In summary, due to the methodology used to determine the maximum

allowable, accident-initiated leak rate (prescribed in Section 2.b.4

of Generic Letter 95-05), the calculated radiological consequences

at the EAB [exclusion area boundary] and LPZ [low population zone]

are larger than previously reported for the postulated steamline

break event. However, the calculated radiological consequences

remain in compliance with NUREG-0800 and GDC-19. Therefore, it is

concluded that the proposed changes do not result in a significant

increase in the radiological consequences of an accident previously

analyzed.

The removal from the FSAR [final safety analysis report] of the

steamline break radiological dose consequences calculation typically

identified as a ``5% failed fuel'' scenario does not affect the

probability or consequences of any accident previously considered.

For CPSES [Comanche Peak Steam Electric Station], no accident-

induced fuel failures are predicted; therefore, consistent with

NUREG-0800, this scenario is not required to be analyzed or

presented in the FSAR.

In summary, because the implementation of the 1.0 volt voltage-

based plugging criteria at Comanche Peak Unit 1 does not adversely

affect steam generator tube integrity and implementation will be

shown to result in acceptable radiological dose consequences, the

proposed Technical Specification change does not result in any

increase in the probability or consequences of an accident

previously evaluated within the Comanche Peak FSAR.

(2) The proposed license amendment does not create the

possibility of a new or different

[[Page 24203]]

kind of accident from any accident previously evaluated.

Implementation of the proposed steam generator tube 1.0 volt

plugging limit does not introduce any significant changes to the

plant design basis. Neither a single or multiple tube rupture event

would be expected in a steam generator in which the plugging limit

has been applied (during all plant conditions).

The bobbin probe voltage-based tube plugging criteria of 1.0

volt is supplemented by: enhanced eddy current inspection guidelines

to provide consistency in voltage normalization, a 100% eddy current

inspection sample size at the tube support plate elevations, and RPC

[rotating pancake coil] inspection requirements for the larger

indications left in service to characterize the principal

degradation as ODSCC. TU Electric will implement a maximum normal

operating condition primary to secondary leakage rate limit of 150

gpd (0.1 gpm--at room temperature) per steam generator to help

preclude the potential for excessive leakage during all plant

conditions. The 150 gpd leakage limit is more restrictive than the

standard operating leakage limit (of 500 gpd) and is intended to

provide additional margin to accommodate a stress corrosion crack

which might grow at a greater than expected rate or unexpectedly

extend outside the thickness of the tube support plate. Leakage

trending capability consistent with EPRI Report TR-04788, ``PWR

Primary-to-Secondary Leak Guidelines'', has been implemented at

Comanche Peak Unit 1.

As steam generator tube integrity upon implementation of the 1.0

volt plugging limit continues to be maintained through in-service

inspection and primary to secondary leakage monitoring, the

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

previously evaluated is not created.

(3) The proposed license amendment does not involve a

significant reduction in margin of safety.

The use of the voltage-based bobbin probe tube support plate

elevation plugging criteria at Comanche Peak Unit 1 maintains steam

generator tube integrity commensurate with the criteria of

Regulatory Guide 1.121. Regulatory Guide 1.121 describes a method

acceptable to the NRC staff for meeting GDCs 14, 15, 31, and 32 by

reducing the probability or the consequences of steam generator tube

rupture. This is accomplished by determining the limiting conditions

of degradation of steam generator tubing, as established by

inservice inspection, for which tubes with unacceptable cracking

should be removed from service. Upon implementation of the proposed

criteria, even under the worst case conditions, the occurrence of

ODSCC at the tube support plate elevations is not expected to lead

to a steam generator tube rupture event during normal or faulted

plant conditions. The end of cycle distribution of crack indications

at the tube support plate elevations is confirmed to result in

acceptable primary to secondary leakage during all plant conditions

and that radiological consequences are not adversely impacted.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Texas at

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

19497, Arlington, TX 76019 Attorney for licensee: George L. Edgar,

Esq., Morgan, Lewis and Bockius, 1800 M Street, NW., Washington, DC

20036.

NRC Section Chief: Robert A. Gramm

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

Date of amendment request: February 16, 1999.

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TS) Sections 3.6, 3.9, and 3.16

and the associated Bases for those sections for Units 1 and 2. The

proposed changes would consolidate the auxiliary feedwater (AFW) cross-

connect requirements by relocating the electrical power requirements

from Section 3.16 to Section 3.6. The proposal also would clarify the

TS with regard to permitting simultaneous entry into certain conditions

of operation on Units 1 and 2.

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:

Criterion 1--Operation of Surry Units 1 and 2 in accordance with

the proposed TS change does not involve a significant increase in

the probability of occurrence or consequences of an accident

previously evaluated.

The proposed TS change is administrative in nature, and station

operations are not being affected. The accidents considered relative

to this proposed TS change are Rupture of Main Steam Pipe, Loss of

All AC Power, and Loss of Feedwater. The probability of occurrence

of these accidents has been previously evaluated to support Surry TS

Amendment 143/140. The NRC reviewed the PSA [probabilistic safety

analysis] basis during issuance of TS Amendment 143/140 and found it

acceptable. The probability of occurrence of these accidents has

been recently reviewed relative to this proposed TS change. It has

been concluded that the proposed TS change is consistent with the

existing analyses and evaluations and, therefore, will not increase

the probability of occurrence of the identified accidents.

The consequences of the accidents identified above were also

previously evaluated to support Surry TS Amendment 143/140. The PSA

considerations included the AFW cross-connect capability, diesel

generator dependencies, various LCO [limiting condition for

operation] time periods, and a HELB [high energy line break] in the

vicinity of the AFW Pumps. The previous evaluation was recently

reviewed relative to this proposed TS change. This review determined

that the proposed TS change is consistent with the design and

licensing bases supporting the existing Technical Specifications.

The proposed TS change is also consistent with the existing analyses

and evaluations, the consequences of which bound any potential

consequences of the proposed TS change. Therefore, the proposed TS

change will not increase the consequences of the identified

accidents.

Criterion 2--The proposed TS change does not create the

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

previously evaluated.

The possibility for a new or different type of accident than any

previously evaluated is not created since the considerations in the

PSA and evaluations performed to support TS Amendment 143/140 are

not changed by the proposed administrative TS change. The proposed

TS change is consistent with the design and licensing bases

supporting the existing Technical Specifications. Furthermore,

station operations and plant equipment are not being affected and,

therefore, the proposed TS change does not create any new failure

modes or accident precursors.

Criterion 3--The proposed TS change does not involve a

significant reduction in a margin of safety.

The proposed administrative change to Surry Technical

Specifications clarifies the requirements (limiting conditions for

operation (LCO) and action statements) relating to the Auxiliary

Feedwater (AFW) cross-connect by relocating the emergency power

source requirements of TSs 3.16.A.8 and 3.16.B.4 to TS 3.6. The

proposed TS change does not alter the current TS requirements or

bases, as well as maintains the Surry licensing and design basis.

The proposed change does not affect either station operations or

plant equipment, hence the availability of equipment for the

mitigation of accidents is not decreased. Furthermore, the

assumptions governing the accident analyses remain unchanged, and

the consequences of the existing analyses and evaluations remain

bounding. This is an administrative change and as such does not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

[[Page 24204]]

Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams,

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

23219.

NRC Section Chief: Richard L. Emch, Jr.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: February 16, 1999.

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TS) Section 4.2 for Units 1 and 2

to relax the surveillance requirements for reactor coolant pump (RCP)

flywheels. The flywheels provide extended reactor coolant flow

coastdown capability if electric power for the RCPs is lost. Currently,

the flywheels are subjected to an inspection program that meets the

requirements of NRC Regulatory Guide 1.14, Revision 1, dated August

1975. The inspections include an ultrasonic examination (UT) of areas

of high stress concentration at the bore and keyway every three years,

and complete UT every 10 years. The proposed change would require only

a 10-year UT, based upon an analysis presented in a Westinghouse

topical report (WCAP-14535A) which has been reviewed and accepted by

NRC staff.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

a. The reduction of the inspection requirements for the reactor

coolant pump flywheels, as generically approved by the NRC and

technically supported by WCAP-14535A, does not significantly

increase the probability of an accident previously evaluated in the

safety analysis report. The results of WCAP-14535A have been

reviewed and evaluated with the technical basis accepted for

referencing in license applications by the NRC in their letter

entitled ``Acceptance for referencing of Topical Report WCAP-14535,

Topical Report on Reactor Coolant Pump Flywheel Inspection

Elimination,'' dated September 12, 1996.

The proposed Technical Specification change reduces the

surveillance requirements (inspection) on the RCP flywheel. There is

no change in the method of plant operation or system design. The

WCAP-14535A report establishes that the proposed change has a

negligible affect on the probability that the flywheel will fail

given that the flywheels received preservice and inservice

examinations as required previously. Therefore, the proposed change

does not increase the probability of occurrence or consequences of

any previously analyzed accident.

b. The proposed change to reduce the inspection requirements for

the RCP flywheels as generically approved by the NRC and supported

by WCAP-14535A does not create the possibility of a new or different

kind of accident from any accident previously evaluated in the

safety analysis report.

The proposed surveillance requirements (inspection) only reduce

the inspection requirements/frequency for the reactor coolant pump

flywheels, and there is no change in the method of plant operation

or system design.

c. The proposed change reducing the inspection of the RCP

flywheels as generically approved by the NRC and supported by WCAP-

14535A, does not impact the accident analysis assumptions or the

basis of any Technical Specification. As previously stated, the

analysis performed in the WCAP-14535A report established that the

affect on flywheel failure probability was negligible given that the

initial preservice and inservice inspections under the current

requirements were performed. Therefore, the proposed change in

surveillance (inspection) frequency does not involve a significant

reduction in the margin of safety.

The analysis provided herein demonstrates that the proposed

amendment to the Surry Technical Specifications does not involve a

significant increase in the probability or consequences of a

previously evaluated accident, does not create the possibility of a

new or different kind of accident, and does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Swem Library, College of

William and Mary, Williamsburg, Virginia 23185.

Attorney for licensee: Donald P. Irwin, Esq., Hunton and Williams,

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

23219.

NRC Section Chief: Richard L. Emch, Jr.

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 amendment request: April 12, 1999 (TSCR 212).

Description of amendment request: The purpose of the proposed

amendments is to update references in the Technical Specifications. The

update is necessary to reflect relocation of referenced information in

the Final Safety Analysis Report.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

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

the proposed amendments will not create a significant increase in

the probability or consequences of an accident previously evaluated.

The proposed amendment corrects references within the Technical

Specification requirements such that they refer to the correct

information in the updated Final Safety Analysis Report (FSAR). The

references changed due to relocation of the information within the

FSAR. The Technical Specification requirements and intent are not

changed. Therefore, these changes are administrative only and do not

change the design or operation of the Point Beach Nuclear Plant

[PBNP]. Operation of PBNP in accordance with the proposed amendments

cannot increase the probability or consequences of an accident

previously evaluated.

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

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

different kind of accident previously evaluated.

The proposed changes are administrative only and therefore do

not materially change any requirements for the design or operation

of PBNP. Therefore, operation in accordance with the proposed

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

accident from any accident previously evaluated.

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

the proposed amendments does not create a significant reduction in a

margin of safety.

The proposed changes are administrative only; correcting

references within the Technical Specification requirements. No

requirement on the operation or design of the facility is being

changed. Therefore, there is no 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: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

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

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

NRC Section Chief: George F. Dick, Jr., Acting.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the

[[Page 24205]]

Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

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

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

Maryland

Date of application for amendments: November 19, 1998.

Brief description of amendments: The amendments revised Technical

Specification 3.7.6 ``Service Water (SRW) System'' to allow operation

of Calvert Cliffs Unit Nos. 1 and 2 with one SRW plate and frame heat

exchanger in a subsystem secured and removing one containment air

cooler from service to enable the affected SRW subsystem to remain

operable.

Date of issuance: April 14, 1999.

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

30 days.

Amendment Nos.: 230 and 206.

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

revised the Technical Specifications.

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

FR 69333). The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated April 14, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

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

Carolina

Date of amendment request: November 1, 1996, as supplemented May

22, 1998, September 14, 1998, January 4, 1999, and March 19, 1999.

Brief description of amendment: The amendment modified the

Technical Specifications for the Brunswick Steam Electric Plant, Units

1 and 2, to extend the Allowed Outage Time for 4.16kV AC balance of

plant buses and the AC electrical power distribution system load group

buses.

Date of issuance: April 15, 1999.

Effective date: April 15, 1999.

Amendment Nos.: 205 and 235.

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

revises the Technical Specifications.

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

FR 6977). The supplemental submittals of May 22, 1998, September 14,

1998, January 4, 1999, and March 19, 1999, contained clarifying

information only, and did not change the initial no significant hazards

consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 15, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam

Electric Plant, Unit No. 2, Darlington County, South Carolina

Date of application for amendment: October 14, 1998.

Brief description of amendment: The amendment modifies the

acceptance criterion for Surveillance Requirement 3.4.14.2 from the

setpoint value of 465 psig to the analytical limit for the residual

heat removal system of 474 psig reactor coolant system pressure.

Date of issuance: April 20, 1999.

Effective date: April 20, 1999.

Amendment No. 182.

Facility Operating License No. DPR-23. Amendment revises the

Technical Specifications.

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

59587).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 20, 1999.

No significant hazards consideration comments received: No

Local Public Document Room location: Hartsville Memorial Library,

147 West College Avenue, Hartsville, South Carolina 29550.

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

County, Michigan

Date of application for amendment: September 3, 1997, as

supplemented March 13, 1998, and March 18, 1999.

Brief description of amendment: The amendment revises the technical

specifications to delete snubber operability requirements, action

requirements for inoperable snubbers, and snubber testing requirements.

The snubber testing requirements have been relocated to the Palisades

Operating Requirements Manual.

Date of issuance: April 13, 1999.

Effective date: April 13, 1999, and shall be implemented within 60

days.

Amendment No.: 185.

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

Technical Specifications.

Date of initial notice in Federal Register: April 8, 1998 (63 FR

17222). The March 18, 1999, submittal requested a 60-day allowance for

implementation of the amendment. This change was within the scope of

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

initial proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 13, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423-3698.

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power

Plant, Unit 1, Monroe County, Michigan

Date of amendment request: July 17, 1998 (Reference NRC-98-0044).

Brief description of amendment: This amendment revises the Enrico

Fermi

[[Page 24206]]

Atomic Power Plant, Unit 1, License to allow possession of a nominal

amount of special nuclear material.

Date of issuance: April 15, 1999.

Effective date: On the date of issuance of this amendment and must

be fully implemented no later than 60-calendar days from the date of

issuance.

Amendment No.: 16.

Facility Operating License No. DPR-9: Amendment revised the License

by adding new Part 2.B.4 to the License.

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

56240). The NRC's related evaluation of the amendment is contained in a

Safety Evaluation dated April 15, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

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

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

Date of application for amendments: March 15, 1999, and

supplemented by letter dated March 17, 1999.

Brief description of amendments: The amendments delete from the

joint Technical Specifications Section 3.3.7, ``Control Room Area

Ventilation System (CRAVS) Actuation Instrumentation,'' and Section

3.3.8, ``Auxiliary Building Filtered Ventilation Exhaust System

(ABFVES) Actuation Instrumentation.'' These surveillance requirements

are not applicable to Catawba because the sections do not reflect the

design of the Catawba units.

Date of issuance: April 8, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days from the date of issuance.

Amendment Nos.: Unit 1--177; Unit 2--169.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

Date of initial notice in FEDERAL REGISTER: March 24, 1999 (64 FR

14274). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 8, 1999.

No significant hazards consideration comments received: No.

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

Black Street, Rock Hill, South Carolina.

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

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

Date of application for amendments: February 18, 1999.

Brief description of amendments: The amendments revise the

Technical Specifications Surveillance Requirement (SR) 3.6.16.1

regarding surveillance of reactor building access openings, SR 3.6.16.3

regarding surveillance of reactor building structural integrity, and

Administrative Controls 5.5.2 regarding the Containment Leakage Rate

Testing Program. The revised requirements would provide scheduling

flexibility without decreasing quality and safety margin.

Date of issuance: April 9, 1999.

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

within 30 days from the date of issuance.

Amendment Nos.: 178--Unit 1; 170--Unit 2.

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

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

11961). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated April 9, 1999.

No significant hazards consideration comments received: No.

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

Black Street, Rock Hill, South Carolina.

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley Power

Station, Unit No. 1, Shippingport, Pennsylvania

Date of application for amendment: November 11, 1998, as

supplemented February 26, 1999.

Brief description of amendment: The amendment modified License

Condition 2.C(9) to allow, on a one-time only, extension of the steam

generator inspection interval in Technical Specification Surveillance

4.4.5.3.b. This will allow the steam generator inspection interval to

coincide with the thirteenth refueling outage or the end of 500

effective full power days, whichever occurs sooner.

Date of issuance: April 16, 1999.

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

days.

Amendment No: 221.

Facility Operating License No. DPR-66. Amendment revised the

License.

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

66593). The February 26, 1999, letter provided additional information

but did not change the initial proposed no significant hazards

consideration determination or expand the amendment request beyond the

scope of the initial notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 16, 1999.

No significant hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, PA 15001.

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas

Nuclear One, Units 1 and 2, Pope County, Arkansas

Date of amendment request: June 28, 1996, as supplemented by

letters dated February 23 and March 15, 1999.

Brief description of amendments: The amendments revise the

Technical Specifications to permit the containment equipment hatch to

be open during handling of irradiated fuel in containment and core

alterations provided that the capability for closure is maintained.

Date of issuance: April 16, 1999.

Effective date: As of the date of issuance, and shall be

implemented within 30 days of issuance.

Amendment Nos.: Unit 1--Amendment No. 195; Unit 2--Amendment No.

203.

Facility Operating License Nos. DPR-51 and NPF-6: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: August 14, 1996 (61 FR

42280). The February 23 and March 15, 1999, letters provided clarifying

information that did not change the scope of the original application

and the initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 16, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment: April 30, 1998.

Brief description of amendment: The amendment revises the single

largest post-accident load capable of being supplied by the diesel

generators and relocates this value to the Bases for Technical

Specification (TS)

[[Page 24207]]

Surveillance 4.8.1.1.2.c.3. TS Surveillance 4.8.1.1.2.c.3 has been

revised to refer to ``the single largest post-accident load'' rather

than a specific numerical value for diesel generator load reject

testing. This change is consistent with the guidance provided in NUREG-

1432 , ``Improved Standard Technical Specifications for Combustion

Engineering Plants.''

Date of issuance: April 21, 1999.

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

30 days from the date of issuance.

Amendment No.: 204.

Facility Operating License No. NPF-6: Amendment revised the

Technical Specifications.

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

56241). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 21, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: November 13, 1997.

Brief description of amendment: The amendment changes the Appendix

A Technical Specifications (TSs) by revising TS 6.8.4.a, Primary

Coolant Sources Outside Containment, to add portions of the containment

vacuum relief and primary sampling systems to the list of systems

included in the Primary Coolant Sources Outside Containment Program.

Date of issuance: April 21, 1999.

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

of issuance, and shall be implemented within 60 days.

Amendment No.: 150.

Facility Operating License No. NPF-38: Amendment revised the

Technical Specifications.

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

FR 9601). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 21, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit 3, Citrus County, Florida

Date of application for amendment: October 1, 1997, as supplemented

April 23 and November 17, 1998 and February 19, 1999.

Brief description of amendment: The changes specify criteria for

evaluating the growth of pit-like intergranular attack steam generator

tube degradation identified in tubes in the ``B'' once-through steam

generator (OTSG). Florida Power Corporation also requested to amend the

Improved Technical Specifications to clarify the date by which the OTSG

inservice inspection results are required to be submitted to the NRC.

Date of issuance: April 8, 1999.

Effective date: April 8, 1999.

Amendment No.: 172.

Facility Operating License No. DPR-72: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 22, 1997 (62 FR

54873). The supplemental letters dated April 23 and November 17, 1998,

and February 19, 1999 did not change the original no significant

hazards determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 8, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit 3, Citrus County, Florida

Date of application for amendment: September 9, 1997, as

supplemented November 7 and 25, 1997, and January 20 and October 30,

1998.

Brief description of amendment: The amendment proposed to revise

the Final Safety Analysis Report (FSAR) analysis of the Makeup System

letdown line failure accident. The revised analysis models the event as

being terminated by manual operator action to isolate the line whereas

the original analysis models an automatic isolation of the break.

Date of issuance: April 13, 1999.

Effective date: April 13, 1999.

Amendment No.: 173.

Facility Operating License No. DPR-72: Amendment approves changes

to the Final Safety Analysis Report.

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

FR 50005). The supplemental letters dated November 7 and 25, 1997,

January 20, 1998, and October 30, 1998, did not change the original

proposed no significant hazards consideration determination, or expand

the scope of the amendment request as originally noticed.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 13, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit 3, Citrus County, Florida

Date of application for amendment: October 30, 1998, as

supplemented April 7, 1999.

Brief description of amendment: Changes the Crystal River Unit 3

Technical Specifications to delete a note regarding the number of

required channels for the Degrees of Subcooling function, and to

subdivide the Core Exit Temperature (Backup) function into two new

functions in Table 3.3.17-1, Post-Accident Monitoring Instrumentation.

Date of issuance: April 20, 1999.

Effective date: April 20, 1999.

Amendment No.: 174.

Facility Operating License No. DPR-72: Amendment revised the

Technical Specifications.

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

2246). The April 7, 1999, supplement did not affect the original no

significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 20, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

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

Plant, Unit No. 2, St. Lucie County, Florida

Date of application for amendment: May 27, 1998, as supplemented

October 9, 1998.

Brief description of amendment: Deletes the requirement for

operability of the safety injection tanks in Mode 4 of reactor

operation.

Date of Issuance: April 8, 1999.

Effective Date: Amendment is effective within 30 days of receipt.

Amendment No.: 100.

[[Page 24208]]

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

Technical Specifications.

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

40556). The October 9, 1998 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 April 8, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of application for amendment: November 25, 1998, as

supplemented February 12, 1999.

Brief description of amendment: The amendment approves the proposed

surveillance Technical Specifications related to the once through steam

generator inservice inspections to be completed during the 13R

refueling outage in fall 1999. Related TS Bases changes are also

included.

Date of issuance: April 13, 1999.

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

30 days.

Amendment No.: 209.

Facility Operating License No. DPR-50. Amendment revised the

Technical Specifications.

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

FR 69342).

The February 12, 1999, submittal modified the request, but did not

affect the initial no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 13, 1999.

No significant hazards consideration comments received: No.

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.

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

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of application for amendment: October 15, 1998, as

supplemented February 3, and February 12, 1999.

Brief description of amendment: The amendment authorizes a revision

to the TMI-1 updated final safety analysis report (UFSAR) for use of

revised atmospheric dispersion factors (X/Q) (obtained by utilizing

recent meteorological data) in determining Chapter 14 postulated

accident analysis radiological dose consequences at Technical

Specification Section 5.1.1 defined exclusion area boundary (EAB) and

low population zone (LPZ).

Date of issuance: April 15, 1999.

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

30 days.

Amendment No.: 210.

Facility Operating License No. DPR-50. Amendment authorizes changes

to the UFSAR.

Date of initial notice in Federal Register: November 18, 1999 (63

FR 64117).

The February 3, and February 12, 1999, letters were within the

scope of the original application and did not change the staff's no

significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 15, 1999.

No significant hazards consideration comments received: No.

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.

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 28, 1998, as

supplemented March 1 and 29, 1999.

Brief description of amendment: The amendment revises Technical

Specification (TS) 2.2.1, ``Limiting Safety System Settings-Reactor

Trip Setpoints,'' to reflect revised loss of normal feedwater flow

analyses.

Date of issuance: April 8, 1999.

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

60 days from the date of issuance.

Amendment No.: 232.

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

Technical Specifications.

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

FR 6701). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 8, 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: January 18, 1999.

Brief description of amendment: The amendment revises Technical

Specification (TS) 3.6.1.2, ``Containment Systems--Containment

Leakage,'' and also revises the related TS bases and Final Safety

Analysis Report sections. The revisions relate to changes in the

secondary containment bypass leakage.

Date of issuance: April 14, 1999.

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

60 days from the date of issuance.

Amendment No.: 234.

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

Technical Specifications.

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

FR 6703). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated April 14, 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: February 10, 1999.

Brief description of amendment: The amendment incorporates

alternative inspection requirements into Technical Specification

Surveillance Requirement 3/4.4.10, ``Structural Integrity,'' for the

reactor coolant pump flywheel.

Date of issuance: April 16, 1999.

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

30 days from the date of issuance.

Amendment No.: 169.

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

Technical Specifications.

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

11964).

[[Page 24209]]

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 16, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie

Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,

Minnesota

Date of application for amendments: February 5, 1999, as

supplemented March 1, 1999.

Brief description of amendments: The amendments revise certain

requirements for repair of defective steam generator tubs specified in

Technical Specification 4.12, ``Steam Generator Tube Surveillance,''

based on the latest revision to a previously approved methodology.

Date of issuance: April 15, 1999.

Effective date: April 15, 1999, with full implementation within 30

days.

Amendment Nos.: 144 Unit 1--135 Unit 2.

Facility Operating License Nos. DPR-42 and DPR-60. Amendments

revised the Technical Specifications.

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

11964). The March 1, 1999, supplement provided corrected Technical

Specification pages. This information was within the scope of the

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

proposed no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 15, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,

Unit No.1, Washington County, Nebraska

Date of amendment request: March 18, 1998.

Brief description of amendment: The amendment revises Technical

Specification (TS) 5.2.f and TS 5.11.2 to change the title of ``Shift

Supervisor'' to ``Shift Manager.''

Date of issuance: April 15, 1999.

Effective date: April 15, 1999.

Amendment No.: 190.

Facility Operating License No. DPR-40. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 8, 1998 (63 FR

17227).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 15, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: May 16, 1996.

Brief description of amendment: The amendment revises requirements

for Plant Operating Review Committee review of fire protection program

and procedure changes.

Date of issuance: April 12, 1999.

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

30 days.

Amendment No.: 252.

Facility Operating License No. DPR-59: Amendment revised the

Technical Specifications.

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

34895).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 12, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

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

York 13126.

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

Generating Station, Salem County, New Jersey

Date of application for amendment: December 16, 1998, as

supplemented March 22, 1999.

Brief description of amendment: This amendment revised Technical

Specification (TS) Surveillance Requirements 4.8.1.1.2 and 4.8.1.1.3,

Table 4.8.1.1.2-1, and the associated Bases. These changes removed the

emergency diesel generator accelerated testing and special reporting

requirements from the TSs in accordance with the guidance provided in

Generic Letter 94-01.

Date of issuance: April 14, 1999.

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

within 60 days.

Amendment No.: 119.

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

Technical Specifications.

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

2251).

The supplemental letters 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 April 14, 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 No. 50-354, Hope Creek

Generating Station, Salem County, New Jersey

Date of application for amendment: June 12, 1998, as supplemented

July 23, 1998 and September 8, 1998.

Brief description of amendment: The amendment revises Technical

Specification (TS) Limiting Condition for Operation Sections 3.7.1.1,

3.7.1.2, and 3.7.1.3. Specifically, the changes revise the Ultimate

Heat Sink limits for river water temperature, in order to increase

operational flexibility. In addition, the Station Service Water System

(SSWS) and Safety Auxiliaries Cooling System (SACS) TS Action

Statements have been revised to provide additional restrictions on

continued plant operation. These revisions provide more explicit TS

direction for plant operation under limiting SSWS/SACS configurations.

Date of issuance: April 19, 1999.

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

within 60 days.

Amendment No.: 120.

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

Technical Specifications.

Date of initial notice in Federal Register: July 1, 1999 (63 FR

35995) The July 23, 1998, and September 8, 1998, supplements provided

clarifying information that did not change the initial proposed no

significant hazards consideration determination or expand the scope of

the original Federal Register notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 19, 1999.

No significant hazards consideration comments received: No

[[Page 24210]]

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070.

South Carolina Electric & Gas Company, South Carolina Public Service

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

No. 1, Fairfield County, South Carolina

Date of application for amendment: September 18, 1998, as

supplemented by letter dated February 5, 1999.

Brief description of amendment: The amendment revises Virgil C.

Summer Nuclear Station Technical Specifications to permit use of the

BEACON system. BEACON is a core power distribution monitoring and

support system based on a three-dimensional nodal code.

Date of issuance: April 9, 1999.

Effective date: April 9, 1999.

Amendment No.: 142.

Facility Operating License No. NPF-12: Amendment revises the

Technical Specifications.

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

FR 64121).

The February 5, 1999, submittal contained clarifying information

only, and did not change the initial no significant hazards

consideration determination. The Commission's related evaluation of the

amendment is contained in a Safety Evaluation dated April 9, 1999.

No significant hazards consideration comments received: No

Local Public Document Room location: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180.

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 application for amendments: January 24, 1997.

Brief description of amendments: The amendments revised

Surveillance Requirement (SR) 3.8.1.9 to Technical Specification 3.8.1,

``AC Sources--Operating,'' to more accurately reflect test conditions

and plant design requirements.

Date of issuance: April 9, 1999.

Effective date: April 9, 1999, to be implemented within 30 days

from the date of issuance.

Amendment Nos.: Unit 2-151; Unit 3-143.

Facility Operating License Nos. NPF-10 and NPF-15: The amendments

revised the Technical Specifications.

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

FR 6997).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 9, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Main Library, University of

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

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: January 20, 1999.

Brief description of amendments: The amendments revise the

descriptive details of Technical Specification 4.7.1.2.1.a, regarding

performance testing of the Auxiliary Feedwater (AFW) pumps, to more

closely adhere to NUREG-1431, ``Improved Standard Technical

Specifications for Westinghouse Plants.'' This involves relocating the

surveillance-required numerical values for the AFW pump performance

test discharge pressure and flow rate to the South Texas Project

Updated Final Safety Analysis Report.

Date of issuance: April 16, 1999.

Effective date: As of the date of issuance and shall be implemented

within 30 days from the date of issuance.

Amendment Nos.: Unit 1--Amendment No. 105; Unit 2--Amendment No.

92.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: February 24, 1999 (64

FR 9201).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 16, 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.

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: January 26, 1999.

Brief description of amendments: The amendments revise part of the

inservice inspection requirements for the reactor coolant pump flywheel

from an in-place ultrasonic volumetric examination of the areas of

higher stress concentration at the bore and keyway at approximately 3-

year intervals and a surface examination of all exposed surfaces and

complete ultrasonic volumetric examination at approximately 10-year

intervals to ultrasonic examination over the volume from the inner bore

of the flywheel to the circle of one-half the outer radius once every

10 years.

Date of issuance: April 16, 1999.

Effective date: April 16, 1999, to be implemented within 30 days of

issuance.

Amendment Nos.: Unit 1--Amendment No. 106; Unit 2--Amendment No.

93.

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

revised the Technical Specifications.

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

11968).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 16, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wharton County Junior College,

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

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: September 30, 1998.

Brief description of amendments: Revises Units 1 and 2 Technical

Specification (TS) Section 3/4.4.5, ``Steam Generator'' Surveillance

Requirements. The future installation of the new Delta 94 steam

generators at the South Texas Project, Units 1 and 2 necessitates

changes to the steam generator tube sample selection and inspection

requirements; inservice inspection frequencies; acceptance criteria;

and inspection reporting requirements.

Date of issuance: April 19, 1999.

Effective date: April 19, 1999, to be implemented following the

replacement of Unit 1 Model E steam generators with Model delta94 steam

generators and prior to Unit 1 operation with the delta94 steam

generators installed.

Amendment Nos.: Unit 1--Amendment No. 107; Unit 2--Amendment No.

94.

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

revised the Technical Specifications.

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

59595).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 19, 1999.

No significant hazards consideration comments received: No.

[[Page 24211]]

Local Public Document Room location: Wharton County Junior College,

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

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: August 6, 1997, as supplemented by

letters dated September 4 and 18, 1997, December 9, 1997, and February

4, 1999.

Brief description of amendments: The amendments revise Technical

Specification (TS) Table 2.2-1 and TS 3/4.2.5 to allow the reactor

coolant system total flow rate to be determined using cold leg elbow

tap differential pressure measurements.

Date of issuance: April 19, 1999.

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

7 days of issuance.

Amendment Nos.: Unit 1--Amendment No. 108; Unit 2--Amendment No.

95.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: August 14, 1997 (62 FR

43556).

The September 4 and 18, 1997, December 9, 1997, and February 4,

1999, letters provided clarifying information that did not change the

original application and the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 19, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Wharton County Junior College,

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

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear Plant,

Unit 1, Hamilton County, Tennessee

Date of application for amendments: August 27, 1998, supplemented

by letter dated March 19, 1999 (TS 98-04).

Brief description of amendments: The amendments change the

Technical Specifications (TS) for Sequoyah Nuclear Plant, Unit 2

reactor by adding a sentence at the end of TS Section 5.3 authorizing

installation of a limited number of lead test assemblies containing

downblended uranium in accordance with Topical Report BAW-2328.

Date of issuance: April 12, 1999.

Effective date: April 12, 1999.

Amendment Nos.: 234.

Facility Operating License No. DPR-79: The amendment revises the

TS.

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

11969). The supplemental letter of March 19, 1999 did not change the

initial proposed no significant hazards condition determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 12, 1999.

No significant hazards consideration comments received: None.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application for amendment: June 29, 1998, as supplemented

by letter dated February 19, 1999.

Brief description of amendment: The amendment revised Technical

Specification (TS) 3.7.1.7 operability requirements to require four

atmospheric steam dump (ASD) lines to be operable. Other changes were

made to TS 3.7.1.7 to address action statements and surveillance

requirements for the four ASD lines.

Date of issuance: April 20, 1999.

Effective date: April 20, 1999, to be implemented within 30 days

from the date of issuance.

Amendment No.: 131.

Facility Operating License No. NPF-30: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 9, 1998 (63

FR 48271).

The February 19, 1999, supplemental letter provided additional

clarifying information, did not expand the scope of the application as

originally noticed, and did not change the staff's original proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 20, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Elmer Ellis Library,

University of Missouri, Columbia, Missouri 65201.

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: May 28, 1998, as supplemented

December 11, 1998.

Brief description of amendments: These amendments revise Technical

Specifications (TS) to provide a specific numerical setting for reactor

trip, reactor coolant pump trip, and auxiliary feedwater initiation on

a loss of power to the 4 kilovolt (kV) buses. Changes to the bases for

the affected TS sections are also being made.

Date of issuance: April 23, 1999.

Effective date: April 23, 1999.

Amendment Nos.: Unit 1-189; Unit 2-194.

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

revised the Technical Specifications.

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

38208).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated April 23, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Dated at Rockville, Maryland, this 28th day of April 1999.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project Management, Office of Nuclear

Reactor Regulation.

[FR Doc. 99-11119 Filed 5-4-99; 8:45 am]

BILLING CODE 7590-01-P