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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory

Commission (the Commission or NRC staff) is publishing this regular

biweekly notice. Public Law 97-415 revised section 189 of the Atomic

Energy Act of 1954, as amended (the Act), to require the Commission to

publish notice of any amendments issued, or proposed to be issued,

under a new provision of section 189 of the Act. This provision grants

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from August 14, 1999, through August 27, 1999.

The last biweekly notice was published on August 25, 1999 (64 FR

46424).

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.

[[Page 48859]]

However, should circumstances change during the notice period such that

failure to act in a timely way would result, for example, in derating

or shutdown of the facility, the Commission may issue the license

amendment before the expiration of the 30-day notice period, provided

that its final determination is that the amendment involves no

significant hazards consideration. The final determination will

consider all public and State comments received before action is taken.

Should the Commission take this action, it will publish in the Federal

Register a notice of issuance and provide for opportunity for a hearing

after issuance. The Commission expects that the need to take this

action will occur very infrequently.

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

Directives Branch, Division of Administration Services, Office of

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

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

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

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

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

written comments received may be examined at the NRC Public Document

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

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

is discussed below.

By October 8, 1999, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

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

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

Date of amendment request: August 3, 1999.

[[Page 48860]]

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 2.1.B to increase the minimum

critical power ratio for higher cycle exposures for Unit 2. The

proposed amendments would also revise TS 6.9.A.6.b for Units 2 and 3 to

add an NRC-approved topical report to the list of analytical

methodologies that are used to determine operating limits.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The probability of an evaluated accident is derived from the

probabilities of the individual precursors to that accident. The

consequences of an evaluated accident are determined by the

operability of plant systems designed to mitigate those

consequences. Limits have been established consistent with NRC-

approved methods to ensure that fuel performance during normal,

transient, and accident conditions is acceptable. These changes do

not affect the operability of plant systems, nor do they compromise

any fuel performance limits.

Changing the Minimum Critical Power Ratio (MCPR) Safety Limit

(SL) at Dresden Nuclear Power Station Unit 2 will not increase the

probability or the consequences of an accident previously evaluated.

This change implements the MCPR SL resulting from the Siemens Power

Corporation (SPC) ANFB critical power correlation methodology using

the approved ATRIUM-9B additive constant uncertainty. For each

cycle, specific MCPR SL calculations will be performed, consistent

with SPC's approved methodology, to confirm the appropriateness of

the MCPR SL. Additionally, operational MCPR limits will be applied

that will ensure the MCPR SL is not violated during all modes of

operation and anticipated operational occurrences. The MCPR SL

ensures that less than 0.1% of the rods in the core are expected to

experience boiling transition. Therefore, the probability or

consequences of an accident will not increase.

Adding EMF-85-74, Revision 0, Supplements 1 and 2 (P)(A) to

Section 6 for Dresden Nuclear Power Station Units 2 and 3, does not

increase the probability or consequences of an accident previously

evaluated. The NRC-approved burnup extension for RODEX2A

applications has been demonstrated to meet all applicable design

criteria. Therefore, adding this methodology to Technical

Specification Section 6 does not increase to 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:

Creation of the possibility of a new or different kind of

accident would require the creation of one or more new precursors of

that accident. New accident precursors may be created by

modifications to the plant configuration, including changes in

allowable modes of operation. This Technical Specification submittal

does not involve any modifications to the plant configuration or

allowable modes of operation. No new precursors of an accident are

created and no new or different kinds of accidents are created.

Therefore, the proposed changes do not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

Changing the MCPR SL does not create the possibility of a new

accident from any accident previously evaluated. This change does

not alter or add any new equipment or change modes of operation. The

MCPR SL is established to ensure that 99.9% of the rods avoid

boiling transition.

The MCPR SL is changing for Dresden Nuclear Power Station Unit 2

to support Cycle 17 operation. This change does not introduce any

physical changes to the plant, the processes used to operate the

plant, or allowable modes of operation. Therefore, no new accidents

are created that are different from any accident previously

evaluated.

The addition of RODEX2A (EMF-85-74, Revision 0, Supplements 1

and 2 (P)(A)) to Section 6 does not create the possibility of a new

accident from an accident previously evaluated. This change does not

alter or add any new equipment or change modes of operation. This

change does not introduce any physical changes to the plant, the

processes used to operate the plant, or allowable modes of

operation. Therefore, no new accidents are created that are

different from any accident previously evaluated.

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

the following reasons:

Changing the MCPR SL for Dresden Nuclear Power Station Unit 2

will not involve any reduction in margin of safety. The MCPR SL

provides a margin of safety by ensuring that less than 0.1% of the

rods are calculated to be in boiling transition. The proposed

Technical Specification amendment request reflects the MCPR SL

results from evaluations by SPC using NRC-approved methodology.

Because the methodology used to determine the MCPR SL is

conservative and has received NRC approval, a decrease in the margin

to safety will not occur due to changing the MCPR SL. The revised

MCPR SL will ensure the appropriate level of fuel protection.

Additionally, operational limits will be established based on the

proposed MCPR SL to ensure that the MCPR SL is not violated during

all applicable modes of operation including anticipated operation

occurrences. This will ensure that the fuel design safety criterion

of more than 99.9% of the fuel rods avoiding transition boiling

during normal operation as well as during an anticipated operational

occurrence is met.

The addition of EMF-85-74, Revision 0, Supplements 1 and 2

(P)(A) to Section 6 does not decrease the margin of safety. The

burnup limit extension for RODEX2A applications has been reviewed

and approved by the NRC. The data supporting the burnup extension

demonstrates that all applicable design criteria are met. Therefore,

since the burnup extension is acceptable and within the design

criteria, using the approved burnup extension will not affect 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: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

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. 50-373 and 50-374, LaSalle

County Station, Units 1 and 2, LaSalle County, Illinois

Date of amendment request: August 13, 1999, as supplemented on

August 27, 1999.

Description of amendment request: The proposed amendments would

revise Technical Specification Section 1.0, ``Definitions,'' Item 1.7,

``Core Alteration,'' to specify that movement of instrumentation and

control rod movements are not considered core alterations if there are

no fuel assemblies in the associated cell. The licensee also proposed

corresponding changes to TS Sections 3/4.1, 3/4.3, and 3/4.9 to reflect

the change in definition.

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 proposed changes incorporate a definition contained in

NUREG-1433, Revision 1, ``Standard Technical Specifications, General

Electric Plants, BWR/4.'' There are no modifications to plant

equipment or systems and there is no direct effect on plant

operation. The proposed changes do not affect any accident

initiators or precursors and do not change or alter the design

assumptions for systems or components used to mitigate the

consequences of an accident. The proposed changes do not affect the

design or operation of any system, structure, or component in the

plant. The proposed changes do not impact

[[Page 48861]]

the requirements for refueling evolutions associated with shutdown

margin, core monitoring, and reactor protection system operability.

There are no changes to parameters governing plant operation, and no

new or different types of equipment will be installed. These changes

do not impact any accident previously evaluated in the Updated Final

Safety Analysis Report (UFSAR). Therefore, no increases in the

probability of an accident or consequences will result due to this

change.

2. Do the proposed changes 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 plant system, structure, or component. There are no changes to

parameters governing plant operation, and no new or different type

of equipment will be installed. There is no change in any method by

which a safety related system performs its function. No new

equipment is being introduced, and installed equipment is not being

operated in a new or different manner. There are no setpoints

affected by this proposed action. This proposed action will not

alter the manner in which equipment operation is initiated, nor will

the function demands on credited equipment be changed. As such, no

new failure modes are being introduced. There are no changes to

assumptions in accident analysis. Therefore, the proposed changes do

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

from any accident previously evaluated.

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

margin of safety?

The proposed changes are consistent with NUREG-1433, Revision 1,

``Standard Technical Specifications, General Electric Plants, BWR/

4.'' The proposed changes do not adversely affect existing plant

safety margins or the reliability of the equipment assumed to

operate in the safety analysis. The initial conditions and

methodologies used in the accident analyses remain unchanged.

Therefore, accident analyses results are not impacted. There are no

resulting effects on plant safety parameters or setpoints. The

proposal does not involve a significant relaxation of the criteria

used to establish safety limits, a significant relaxation of the

bases for the limiting safety system settings, or a significant

relaxation of the bases for the limiting conditions for operations.

Therefore, these proposed changes do not cause a 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: Jacobs Memorial Library, 815

North Orlando Smith Avenue, Illinois Valley Community College, Oglesby,

Illinois 61348-9692.

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.

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

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

Date of amendment request: May 5, 1999.

Description of amendment request: The proposed amendment would

permit a one-time extension of the allowed outage time (AOT) for the

reactor protection and engineered safety feature actuation

instrumentation.

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 reactor protection and engineered safety features functions

are not initiators of any design basis accident or event and

therefore do not increase the probability of any accident previously

evaluated. The proposed changes to the AOTs, bypass times, and

allowing on-line testing and maintenance have an insignificant

impact on plant safety based on the calculated CDF [core damage

frequency] increase being less than LOE-06. Therefore, the proposed

changes do not result in a significant increase in the consequences

of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes do not result in a change in the manner in

which the RPS [reactor protection system] and ESFAS [engineered

safety features actuation system] provide plant protection. No

change is being made which alters the functioning of the RPS and

ESFAS. Rather, the likelihood or probability of the RPS or ESF

functioning properly is affected as described above. Therefore, the

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

kind of accident nor involve a reduction in the margin of safety as

defined in the Safety Analysis Report.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes do not alter the manner in which safety

limits, limiting safety system setpoints or limiting conditions for

operations are determined. The impact of increased AOTs, testing

times, and allowing on-line testing and maintenance are expected to

result in an overall improvement in safety because:

The longer AOTs for the master relays, logic cabinets, and

analog channels will promote improved maintenance practices that

will provide improved component performance, improved availability

of the protection system, and a reduced number of spurious reactor

trips and spurious actuation of safety equipment.

The longer AOTs and bypass times for the analog channels will

provide additional time before being required to place the channel

in trip. With the channel in trip, the logic required to cause a

reactor trip or a safety system actuation is reduced to 1 of 2 (for

2 of 3 logic) and to 1 of 3 (for 2 of 4 logic). With the reduced

logic requirement, the potential for a spurious actuation is

increased. Leaving the channel in the bypass state for additional

time does reduce the availability of signals to initiate component

actuation for event mitigation when required, but as shown in this

analysis, the impact on plant safety is small due to the

availability of other signals or operator action to trip the reactor

or cause component actuation.

The longer allowed outage times will provide plant operators

additional flexibility in operating the plant. There will be

additional time available before an action needs to be taken to shut

down the plant or place a channel in the tripped state. This

additional flexibility will facilitate prioritizing component

repairs.

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: White Plains Public Library,

100 Martine Avenue, White Plains, New York 10610. Biweekly Notice

Coordinator Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving

Place, New York, New York 10003.

NRC Section Chief: S. Singh Bajwa.

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: August 4, 1999.

Description of amendment request: The amendments would revise the

joint Technical Specifications as follows:

(1) A current action in Section 3.2.2 requires that when one

Nuclear Service Water System (NSWS) suction transfer low pit level

channel is inoperable, the channel be placed in its trip position. The

licensee proposed an additional alternative such that the NSWS suction

can simply be aligned from Lake Wylie to the Standby Nuclear Service

Water Pond (SNSWP). Suction from Lake Wylie is the normal

configuration, while suction from the SNSWP is the safety

configuration. This proposed alternative

[[Page 48862]]

action provides operational flexibility; there is no associated design

change to the units.

(2) The licensee proposed to delete from Table 3.3.2-1,

``Engineered Safety Feature Actuation System Instrumentation,'' the

entry regarding Auxiliary Feedwater Loss of Offsite Power (Function 6d)

on the basis that a comparable and adequate requirement will exist in

Section 3.3.5. To such end, a new Surveillance Requirement (SR) 3.3.5.3

will be added, incorporating the Function 6d requirement from Table

3.3.2-1. These proposed changes remove inconsistencies that currently

exist in the Technical Specifications for Function 6d. There is no

associated design change to the units.

(3) In the process of converting the Technical Specification to the

improved format (Amendment Nos.173 and 165), errors were inadvertently

introduced regarding the conditions under which the Reactor Coolant

System Subcooling Margin Monitor must be operable. The licensee

proposed to correct these errors by revising the entry regarding the

Subcooling Margin Monitor in Table 3.3.3-1, ``Post Accident Monitoring

Instrumentation''. There is no associated design change to the units.

(4) Section 3.4.17 is concerned with reactor coolant system loops

test exceptions. Currently Surveillance Requirement 3.4.17.2

incorrectly specifies that a COT [channel operational test] be

performed ``for each power range neutron flux-flow and intermediate

range neutron flux channel and P-7 [Low Power Reactor Trips Block

Function]''. The licensee proposed to correct this statement by

deleting ``P-7'' and adding ``P-10 [Power Range Neutron Flux] and P-13

[Turbine Impulse Pressure]''. This correction does not involve any

design change to the units.

(5) The licensee proposed to delete from Section 5.3.1 the specific

qualification requirements for Reactor Operators (ROs) and Senior

Reactor Operators (SROs). Such requirements are specified by 10 CFR

50.55, ``Operators'' Licenses'', and the licensee is required to follow

this regulation. There will be no change in the qualification of ROs

and SROs, and no design change to the units.

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:

First Standard

Implementation of this amendment would not involve a significant

increase in the probability or consequences of an accident

previously evaluated. Approval of this amendment will have no effect

on accident probabilities or consequences. For proposed changes #1-

4, the systems and equipment referenced in the revised TS are not

accident initiating systems; therefore, there will be no impact on

any accident probabilities by the approval of this amendment. The

design of the systems is not being modified by these proposed

changes. Therefore, there will be no impact on any accident

consequences. For proposed change #5, the change is purely

administrative; it will therefore have no effect on any accident

probabilities or consequences.

Second Standard

Implementation of this amendment would not create the

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

previously evaluated. No new accident causal mechanisms are created

as a result of NRC approval of this amendment request. No changes

are being made to the plant which will introduce any new accident

causal mechanisms. This amendment request does not impact any plant

systems that are accident initiators; neither does it adversely

impact any accident mitigating systems.

Third Standard

Implementation of this amendment would not involve a significant

reduction in a margin of safety. Margin of safety is related to the

confidence in the ability of the fission product barriers to perform

their design functions during and following an accident situation.

These barriers include the fuel cladding, the reactor coolant

system, and the containment system. The performance of these fission

product barriers will not be impacted by implementation of this

proposed amendment. The systems and equipment referenced in the

revised TS for proposed changes #1-4 are already capable of

performing as designed. No safety margins will be impacted. Since

proposed change #5 is purely administrative, it will have no effect

on any safety margins.

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.

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

Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: July 26, 1999.

Description of amendment request: The proposed amendment would

change Technical Specification (TS) Section 3/4.3.2.1, ``Safety

Features Actuation System Instrumentation,'' to remove the ``Trip

Setpoint'' values and revise the ``Allowable Values'' entries for

Sequence Logic Channels a, ``Essential Bus Feeder Breaker Trip (90%),''

and b, ``Diesel Generator Start, Load Shed on Essential Bus (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:

The Davis-Besse Nuclear Power Station (DBNPS) has reviewed the

proposed changes and determined that a significant hazards

consideration does not exist because operation of the Davis-Besse

Nuclear Power Station, Unit No. 1, in accordance with these changes

would:

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

accident previously evaluated because the proposed changes do not

change any accident initiator, initiating condition, or assumption.

The proposed changes would revise Technical Specification (TS)

Table 3.3-4, Safety Features Actuation System Instrumentation Trip

Setpoints, to remove the'Trip Setpoint'' values for Functional Unit

Sequence Logic Channel ``a'', ``Essential Bus Feeder Breaker Trip

(90%)'', and Functional Unit Sequence Logic Channel ``b'', ``Diesel

Generator Start, Load Shed on Essential Bus (59%)'', and also modify

the ``Allowable Values'' entry for Functional Unit Sequence Logic

Channel ``a'', consistent with updated calculations and current

setpoint methodology. The proposed changes would also clarify an

inconsistency between Table 3.3-4 and Table 4.3-2, Safety Features

Actuation System Instrumentation Surveillance Requirements. The

proposed changes to Limiting Condition for Operation (LCO) 3.3.2.1

and Bases 3/4.3.1 and 3/4.3.2 are associated with these changes.

The accident previously evaluated in Section 15.2.9, ``Loss of

All AC Power to the Station Auxiliaries (Station Blackout),'' of the

DBNPS Updated Safety Analysis Report (USAR) is not affected by the

proposed changes because its bounding conditions are not affected.

The existing TS action statements will continue to maintain the USAR

requirement to start and load one Emergency Diesel Generator (EDG)

to meet minimum ESF requirements, should all AC power be lost.

Furthermore, the proposed changes are based on the existing

performance characteristics of plant equipment; therefore, the

proposed changes

[[Page 48863]]

will not involve a significant change to the plant design or

operation.

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

accident previously evaluated because the proposed changes do not

invalidate assumptions used in evaluating the radiological

consequences of an accident, do not alter the source term or

containment isolation, and do not provide a new radiation release

path or alter radiological consequences.

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

accident from any accident previously evaluated because the proposed

changes do not introduce a new or different accident initiator or

introduce a new or different equipment failure mode or mechanism.

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

because the proposed changes do not significantly reduce the ability

of the plant to respond to a loss of AC power to the essential 4160

Volt buses in a timely manner. The revised Allowable Value for the

Sequence Logic Channel ``Essential Bus Feeder Breaker Trip (90%)''

takes into account the need not only to be able to actuate

Engineered Safety Features equipment coincident with a degraded grid

condition, but to provide voltage at the required value to properly

operate the equipment.

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.

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

Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: July 27, 1999.

Description of amendment request: The proposed amendment would

remove Technical Specification (TS) Section 6.4, ``Training,'' relocate

TS Sections 6.5.2.8, ``Audits,'' and 6.10 ``Record Retention,'' to the

Updated Safety Analysis Report, and make related changes to TS Sections

6.14, ``Process Control Program,'' and 6.15, ``Offsite Dose Calculation

Manual.'' In addition, an editorial correction is proposed to TS 6.8,

``Procedures and Programs.''

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 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, Unit Number 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

Section 6.0, Administrative Controls, of the Technical

Specifications (TS).

The proposed changes to remove Section 6.4, Training, from the

TS and relocate the detailed listings of TS Section 6.5.2.8, Audits,

and TS Section 6.10, Record Retention, to the DBNPS [Davis-Besse

Nuclear Power Station] Quality Assurance Program in Chapter 17 of

the Updated Safety Analysis Report are consistent with NUREG-1430,

``Standard Technical Specifications--Babcock and Wilcox Plants,''

Revision 1 or NRC Administrative Letter 95-06 ``Relocation of

Technical Specification Administrative Controls Related to Quality

Assurance,'' dated December 12, 1995. The proposed changes to TS

Section 6.14, Process Control Program (PCP); TS Section 6.15,

Offsite Dose Calculation Manual (ODCM); and TS Section 6.8,

Procedures and Programs, are either associated administratively with

the above proposed changes or are editorial corrections. These TS

being removed or relocated will remain subject to the controls of

regulations (e.g., 10 CFR 50.59, 10 CFR 55.59, or 10 CFR 50.54(a)).

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

accident previously evaluated because no accident conditions or

assumptions are affected by the proposed changes. As described

above, these changes are consistent with the improved ``Standard

Technical Specifications--Babcock and Wilcox Plants'' (NUREG-1430)

or Administrative Letter 95-06 and are administrative changes. The

proposed changes do not alter the source term, containment

isolation, or allowable releases. The proposed changes, therefore,

will not increase the radiological consequences of a previously

evaluated accident.

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

accident from any accident previously evaluated because no new

accident initiators or assumptions are introduced by the proposed

changes, which involve only administrative controls. The proposed

changes do not alter any accident scenarios.

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

because the proposed changes are administrative and do not reduce or

adversely affect the capabilities of any plant structures, systems

or components to perform their nuclear safety function.

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.

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

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

Date of amendment request: August 5, 1999.

Description of amendment request: The requested changes correct

editorial errors in Technical Specification (TS) Sections 3.8.3.2,

4.6.2.1, 4.6.2.2, 4.8.1.1, and 4.9.12. Also, the requested changes

correct minor editorial and reference errors in Technical Specification

Bases Sections B 3/4.3.2, B 3/4.4.11, B 3/4.6.1.2, and B 3/4.8.4.

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 [Northeast Nuclear Energy Company] has reviewed the

proposed revision in accordance with 10CFR50.92 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 changes are editorial in nature and do not alter

or effect the design, operation, maintenance[,] or surveillance

associated with MP-3 [Millstone Nuclear Power Station, Unit No. 3]

[s]tructures, [s]ystems, and [c]omponents (SSC) during normal or

accident operations. Since the SS[Cs] are not altered[,] the

proposed changes do not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

from any accident previously evaluated.

The proposed TS changes are editorial in nature and do not alter

or effect the design,

[[Page 48864]]

operation, maintenance[,] or surveillance associated with MP-3

[s]tructures, [s]ystems, and [c]omponents (SSC) during normal or

accident operations. Since the Units SS[Cs] have not been modified

physically, or operationally[,] due to procedure changes prompted by

this TSCR [Technical Specification Change Request], the proposed

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

accident from any previously evaluated.

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

These proposed TS changes are editorial and do not impact any

MP-3 design or operational requirements. MP-3 system performance and

operating limits are not affected; therefore[,] the proposed change

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

In conclusion, based on the information provided, it is

determined [by NNECO] that the proposed revision does not involve

a[n] 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, Docket Nos. 50-352 and 50-353, Limerick Generating

Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of amendment request: June 22, 1999.

Description of amendment request: The Limerick Generating Station

(LGS), Units 1 and 2, Technical Specifications (TS) contained in

Appendix A to the Operating Licenses would be amended to eliminate a

surveillance requirement for the Reactor Recirculation System. This

proposed TS change request involves revising the TS to delete

Surveillance Requirement 4.4.1.1.2, and associated TS Administrative

Controls Section 6.9.1.9.h, which requires that each Reactor

Recirculation System pump motor generator (MG) set scoop tube

mechanical and electrical stop be demonstrated OPERABLE with the

overspeed setpoints less than or equal to the setpoints as noted in the

Core Operating Limits 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. 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 Technical Specifications (TS) changes do not

involve a significant increase in the probability or consequences of an

accident previously evaluated. The proposed TS changes do not make any

physical changes to the fuel, or the way the fuel responds to a

transient or accident. The radiological barriers are not compromised.

The fuel will continue to be operated to analyzed operating limits. No

new failure mode is introduced.

Prior to the removal of the Recirculation System Master Flow

Controller at LGS, the bounding postulated event involving an increase

in reactor coolant system flow rate was the dual pump slow flow runout

event not terminated by SCRAM. The requirements surrounding the MG set

stops were established to mitigate consequences during a dual pump slow

flow runout by providing a limit on the maximum core flow. The MG set

stop requirements were not established to prevent an accident. The

potential common mode failure required for a dual pump slow flow runout

event was eliminated with the removal of the Master Flow Controller.

The elimination of the Master Flow Controller does not increase the

probability of other core flow increase events, or of any other events

previously analyzed.

Revised generic flow biased ARTS [APRM (average power range

monitor)/RBM (rod block monitor) Technical Specifications Improvement]

thermal limits that do not take credit for MG set stops have been

developed for LGS, Units 1 and 2. Adherence to approved flow biased

ARTS thermal limits identified in the LGS, Units 1 and 2, Core

Operating Limits Reports (COLRs) ensure that fuel design limits are not

exceeded. Maintaining fuel design limits results in no change in the

consequences of accidents previously evaluated.

The single pump slow flow runout does not terminate by Main Steam

Isolation Valve (MSIV) closure or generator load reject. As a result,

the single pump runout event does not result in any significant

pressurization and does not represent a challenge to the reactor

coolant pressure boundary. MSIV closure with associated SCRAM on high

neutron flux, as confirmed in the cycle specific Supplemental Reload

Licensing Report (SRLR), remains the bounding reactor pressure vessel

overpressurization event for LGS, Units 1 and 2. In addition, there are

no other associated impacts to the plant resulting from a single pump

runout. Therefore, the integrity of radiological barriers will not be

compromised.

Although there is no longer a safety need to demonstrate

operability of the MG set stops, there still is an operational need to

have the MG set stops for the Reactor Recirculation System (RS). Damage

to the jet pump sensing lines could occur if the resonance frequency of

the sensing lines is reached. Jet pump sensing line tests established a

conservative pump speed limit (1650 rpm for Unit 1, no limit for Unit

2) to preclude sensing line resonance. The MG set stop setpoint bounded

the operationally required setpoint. The operationally required MG set

stop setpoint to preclude jet pump sensing line resonance will continue

to be controlled administratively via approved plant procedures. The

proposed TS changes do not adversely impact the RS, or introduce new or

unanalyzed operating conditions for the RS. The MG sets will not exceed

their previously analyzed maximum 57.5 Hz with the stops removed.

Therefore, the proposed TS changes do not significantly increase

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously evaluated.

The proposed TS changes do not make any physical changes to the fuel,

or the way the fuel responds to a transient or accident. The

radiological barriers are not compromised. The fuel will continue to be

operated to analyzed operating limits. No new failure mode is

introduced.

The proposed TS changes do not create new operating conditions that

have not been evaluated. Removal of the Recirculation Master Flow

Controller eliminates the possibility of a single failure initiated

common mode event. Since the possibility of a common failure has been

eliminated, the most limiting recirculation runout event is a one pump

slow flow runout. This is the same kind of postulated accident as that

previously evaluated, only it involves one pump instead of both pumps.

Therefore, the proposed TS changes do not create the possibility of a

new or different kind of accident from any previously evaluated.

[[Page 48865]]

3. The proposed TS changes do not involve a significant reduction

in a margin of safety.

The proposed TS changes do not make any physical changes to the

fuel, or the way the fuel responds to a transient or accident. The

radiological barriers are not compromised. The fuel will continue to be

operated to analyzed operating limits. No new failure mode is

introduced.

Single pump runout based, generic flow biased ARTS thermal limits

that do not take credit for MG set stops have been developed for LGS,

Units 1 and 2. Adherence to approved ARTS-based flow biased thermal

limits identified in the LGS, Units 1 and 2, COLRs and implemented in

the plant process computer are sufficient to maintain the margin of

safety as delineated in TS Sections 3/4.2.1, 3/4.2.3, and 3/4.2.4.

Therefore, these proposed TS changes do not involve a significant

reduction in a margin of safety.

Based on the above review, the NRC staff concludes that 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: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

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.

Portland General Electric Company, et al., Docket No. 50-344, Trojan

Nuclear Plant, Columbia County, Oregon

Date of amendment request: January 29, 1998.

Description of amendment request: The amendment would delete the

requirements for a security plan from the 10 CFR Part 50 license and

technical specifications after the spent nuclear fuel is transferred to

a Part 72 licensed independent spent fuel storage installation (ISFSI).

Security requirements for the ISFSI would be in accordance with 10 CFR

Part 72, Subpart H.

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 physical structures, systems and components of the Trojan

Nuclear Plant and the operating procedures for their use are

unaffected by the proposed change. The proposed elimination of the

security requirements for the 10 CFR Part 50 license, is predicated

on approval of the Trojan ISFSI Security Plan (PGE 1073) which will

be coincident with issuance of a 10 CFR Part 72 license and upon

completion of the transfer of all nuclear fuel from the spent fuel

pool to the ISFSI. The planned 10 CFR 72 licensing controls for the

ISFSI will provide adequate confidence that personnel and equipment

can perform satisfactorily for normal operations of the ISFSI and

respond adequately to abnormal events/accidents. The proposed Trojan

ISFSI Security Plan (PGE 1073) will also provide confidence that

security personnel and safeguards systems will perform

satisfactorily to ensure adequate protection for the storage of

spent nuclear fuel. Therefore, the proposed 10 CFR Part 50 amendment

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.

The proposed change is security related, and as such, has no

direct impact on plant equipment or the procedures for operating

plant equipment and, therefore, does not create the possibility of a

new or different kind of accident from any accident previously

evaluated. Because the proposed ISFSI area will be segregated from

the 10 CFR Part 50 licensed area, licensed security activities under

the 10 CFR Part 50 license will no longer be necessary after all the

nuclear fuel has been moved. The planned 10 CFR 72 licensing

controls for the ISFSI area will provide adequate confidence that

personnel and equipment can perform satisfactorily for normal

operations of the ISFSI and respond adequately to normal events/

accidents. Moreover, the ISFSI will be physically separate from the

Trojan Nuclear Plant structures and equipment. Therefore, the

proposed 10 CFR Part 50 license amendment does not create the

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

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The assumptions for a fuel handling and other accidents are not

affected by the proposed license amendment. Because the proposed

ISFSI area (that will contain the nuclear fuel) will be segregated

from the 10 CFR Part 50 licensed area, licensed security activities

under the 10 CFR Part 50 license will no longer be necessary. The

planned 10 CFR 72 licensing controls for the ISFSI area will provide

adequate confidence that personnel and equipment can perform

satisfactorily for normal operations of the ISFSI and respond

adequately to abnormal events/accidents. Also, the ISFSI will be

physically separate from the Trojan Nuclear Plant structures and

equipment. Therefore, the proposed 10 CFR Part 50 license amendment

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: Branford Price Millar Library,

Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,

Portland, Oregon 97207.

Attorney for licensee: Leonard A. Girard, Esq., Portland General

Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.

NRC Section Chief: Michael T. Masnik.

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

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

Unit No. 1, Fairfield County, South Carolina

Date of amendment request: August 19, 1999. The August 19, 1999,

submittal supersedes the February 18, 1999, submittal in its entirety

(64 FR 14284).

Description of amendment request: The proposed amendment would

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

Specifications (TS) to incorporate the new Pressure/Temperature (P-T)

Limits Curves consistent with the analysis results of reactor vessel

specimen W. These figures are contained in Section 3/4.4.9 and are

presented as Figures 3.4-2 and 3.4-3. These figures were developed

using the methodology included in WCAP 14040-NP-A, ``Methodology Used

to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup

and Cooldown Limit Curves,'' as well as Code Case N-640, ``Alternative

Reference Fracture Toughness for Development of P-T Limit Curves for

Section XI, Division I.'' A reduced flange temperature requirement was

included in the development of the curves, with justification provided

in WCAP 15102, Revision 1, ``V. C. Summer Unit I Heatup and Cooldown

Limit Curves for Normal Operation.'' Additionally, the Bases section

for the Pressure/Temperature Limits would be revised to accurately

reflect current industry standards and regulations. A significant

portion of this Bases section would be deleted due to the information

also being located in WCAP 15102, Revision 1.

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

[[Page 48866]]

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes revise the Pressure/Temperature Limits

Curves to provide curves that reflect the results of the analysis

performed on reactor vessel surveillance specimen W. This analysis

was performed using NRC approved methodology as documented in WCAP

14040-NP-A, utilizing the 1996 ASME Boiler and Pressure Vessel Code,

Section XI, Appendix G requirements, along with ASME Code Case N-

640. These curves provide the limits for operation of the Reactor

Coolant System during heat up, cool down, criticality, and

hydrotesting. These curves are provided without instrument

uncertainties included, however, the uncertainties are included in

the curves provided in the operating procedures. The limits protect

the reactor vessel from brittle fracture by separating the region of

acceptable operation from the region where brittle fracture is

postulated to occur. Failure of the reactor vessel is not a VCSNS

design basis accident, and, in general, reactor vessel failure has a

low probability of occurrence and is not considered in the safety

analysis. Therefore, the change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed changes revise the Pressure/Temperature Limits

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

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

changes or significant changes in any operating procedures. This

change adjusts the heatup and cooldown curves to reflect the shift

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

result of neutron embrittlement, and alternate methodology utilized

to generate the curves. Therefore, the change does not create the

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

previously evaluated.

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

safety?

The proposed changes revise the Pressure/Temperature Limits

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

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

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

normal operation including Reactor Coolant System transients during

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

were prepared, using accepted industry methodology, for a projected

reactor vessel neutron exposure of 32 EFPY [Effective Full Power

Years].

The new curves will serve as the basis for operating

limitations, to provide margin against non-ductile fractures. The

uncertainties introduced by instrumentation, forced flow and

elevation differences are not reflected in the TS curves. These

uncertainties will be factored into the curves presented in the

operating procedures. Since administrative limits remain in place to

ensure that 10 CFR 50 Appendix G limits are not challenged, the

margin of safety described in the TS Bases is not reduced by the

proposed change. Therefore, the change does not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Fairfield County Library, 300

Washington Street, Winnsboro, SC 29180.

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

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

NRC Section Chief: Richard L. Emch, Jr.

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

362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego

County, California

Date of amendment requests: August 11, 1999 (PCN-488).

Description of amendment requests: The proposed amendments would

modify the Technical Specifications for the San Onofre Nuclear

Generating Station (SONGS) Units 2 and 3 to revise Surveillance

Requirement (SR) 3.3.7.3 by providing allowable values in place of

analytical limits for certain degraded voltage parameters, and by

deleting unnecessary parameter limits in cases where plant safety is

not affected. The proposed change would also delete redundant SR

3.3.7.4.

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 proposed amendment involve a significant increase in

the probability or consequences of an accident previously evaluated?

No.

Proposed Change Number (PCN)-488 revises the Technical

Specification (TS) Surveillance Requirement (SR) acceptance criteria

of the Loss of Voltage Signal (LOVS), Degraded Grid Voltage with

Safety Injection Actuation Signal (DGVSS), and Sustained Degraded

Voltage Signal (SDVS) relay circuits. These circuits are not

accident initiators.

PCN-488 revises the TS SR acceptance requirements to make them

more limiting than the present requirements. Because the revised

acceptance criteria are more limiting than the present requirements,

the consequences of accidents analyzed in the Updated Final Safety

Analysis Report (UFSAR) are not increased. PCN-488 also revises the

TS SR acceptance requirements to delete upper and lower bounds in

cases where the deleted bound provides no safety benefit. Deleting

bounds having no safety significance does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

PCN-488 deletes redundant SR 3.3.7.4, which is not in NUREG-

1432, Standard Technical Specifications, Combustion Engineering

Plants. Deleting a redundant requirement does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

Consequently, the proposed amendment does not result in an

increase in the probability of accidents evaluated in the UFSAR.

2. Does this amendment request create the possibility of a new

or different kind of accident from any accident previously

evaluated?

No.

PCN-488 revises the TS SR acceptance criteria of the LOVS,

DGVSS, and SDVS relay circuits, which are not accident initiators,

and deletes a redundant SR. PCN-488 does not introduce any revision

in the hardware configuration of the protective circuitry for LOVS,

DGVSS or SDVS. The measurement required by the deleted, redundant

surveillance is required elsewhere in the TS. For these reasons,

PCN-488 does not create the possibility of any new or different kind

of accident from any previously evaluated. '

3. Does this amendment request involve a significant reduction

in a margin of safety?

No.

PCN-488 provides allowable values for the acceptance criteria

for the TS SR for LOVS, DGVSS and SDVS. As such, the revised values

are more limiting than the current values, which represent design

limits. Therefore, PCN-488 does not involve a significant reduction

in a margin of safety.

PCN-488 also revises the TS SR acceptance requirements to delete

upper and lower bounds in cases where the deleted bound provides no

safety benefit. Deleting bounds having no safety significance does

not involve a significant reduction in a margin of safety.

PCN-488 additionally deletes a redundant SR. Because the deleted

surveillance is required elsewhere in the TS, this action does not

involve a significant reduction in a margin of safety.

For these reasons, PCN-488 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 requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

California, Irvine, California 92713.

[[Page 48867]]

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

California Edison Company, 2244 Walnut Grove Avenue, Rosemead,

California 91770.

NRC Section Chief: Stephen Dembek.

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 31, 1998, as supplemented by

letters dated April 19 and August 18, 1999. The August 31, 1998,

application was originally noticed in the Federal Register on October

21, 1998 (63 FR 56260).

Description of amendment request: The proposed amendments would

revise Technical Specification 3/4.4.9.3 by revising the cold

overpressure mitigation curve to accommodate the replacement steam

generators and by adding two surveillances (for the centrifugal

charging pumps and the emergency core cooling system accumulators) to

ensure the operability of the cold overpressure mitigation system.

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.

Reanalysis of STP [South Texas Project, Units 1 and 2] COMS

[cold overpressure mitigation system] transients to consider design

characteristics of Delta-94 RSGs [replacement steam generators] has

shown that maximum allowable PORV [power-operated relief valve]

setpoints decrease slightly, and continue to provide design basis

low temperature overpressure protection with Delta-94 steam

generators. This change request incorporates the new COMS curves

into Technical Specification 3.4.9.3 (Figure 3.4-4). Maximum

allowable PORV setpoints decrease with Delta-94 steam generators,

and are conservative compared to Model E steam generator curves. Use

of the new curves with either Model E or Delta-94 steam generators

conforms to the STP design basis.

These changes are based on a reanalysis that accounts for Model

Delta-94 design, a decision to make calculation[s] of COMS maximum

allowable PORV setpoint consistent with current industry standards

as represented by WCAP-14040, and addition of two surveillances to

the Technical Specification to ensure operability of COMS. Moving

maximum allowable PORV setpoints in the conservative direction and

adding surveillances to reinforce standard operating practice have

no adverse effect on the probability or consequences of an accident

previously evaluated. Therefore, the proposed 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.

The proposed PORV maximum allowable setpoint changes do not

create any new operating conditions or modes, and the added

surveillances have no effect except to ensure operation of COMS as

designed. The slight change to the maximum allowable PORV setpoint

curves for the Cold Overpressure Mitigation System accommodates

Delta-94 steam generator design characteristics, and COMS continues

to perform in accordance with existing requirements, which are

sufficient to ensure plant safety is preserved.

The proposed change is the result of a reanalysis of a

previously evaluated accident. Therefore, the proposed change does

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

from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change reflects design characteristics of the new

Delta-94 steam generators. The change to the COMS curves is in the

conservative direction and does not affect any design failure point

or system limitation. Therefore, the change does not involve a

significant reduction in a margin of safety.

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

this review, it appears that the 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.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: August 18, 1999.

Description of amendment request: The licensee proposed changing

the Vermont Yankee Nuclear Power Station (VY) Technical Specifications

by revising the reactor core spiral reloading pattern such that it

begins around a source range monitor rather than from the center of the

core. The offloading pattern would be the reverse sequence.

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 operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

VY has determined that the proposed change to reload the reactor

core in a spiral pattern beginning around a Source Range Monitor

(SRM) does not involve a significant increase in the probability or

consequences of an accident previously evaluated. The design basis

accident associated with refueling is the Refueling Accident; i.e.,

the accidental dropping of a fuel bundle onto the top of the core.

There is no assumption as to the core loading pattern in the

analysis of this accident. The analyzed abnormal operational

transients associated with refueling are: (1) the Control Rod

Removal Error During Refueling, and (2) the Fuel Assembly Insertion

Error During Refueling. There is no assumption as to the core

loading pattern in the analyses of these transients. The Fuel

Assembly Insertion Error During Refueling transient involves

mislocated and rotated fuel assembly loading errors. However, a

change in the approved core loading pattern has no impact on the

probability of mislocating or rotating a bundle while following that

pattern. Furthermore, the proposed change implements a core loading

pattern that provides improved flux monitoring as compared to the

pattern prescribed by the current Technical Specifications. When

loading the core in accordance with the proposed change, the SRM

indication will be indicative of the true flux of the loaded fuel,

as the creation of flux traps (moderator filled cavities surrounded

on all sides by fuel) is precluded.

The SRMs and the core loading pattern are not initiators of any

accident previously evaluated. As such, the subject changes cannot

affect the probability of an accident previously evaluated. The core

loading pattern is not assumed in the mitigation of any accident.

Since the proposed change provides improved flux monitoring by the

SRMs, operators will have more accurate indication and SRM automatic

trip functions will actuate based on a more accurate indication of

flux. As such, any event mitigation function provided by the SRMs is

enhanced by this change. Therefore, the associated changes do not

involve a significant increase in the consequences of an accident

previously evaluated.

2. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not create the

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

previously evaluated.

VY has determined that the proposed change does not create the

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

previously evaluated. VY proposes to change the core reloading and

offloading patterns to start and stop, respectively, at an

[[Page 48868]]

SRM versus the geometric center of the core as prescribed by current

Technical Specifications. This ensures that flux monitoring

instrumentation is always OPERABLE in the fueled region of the

vessel. There is no separation of the monitoring device from the

fuel by cavities of water as is the case with the pattern prescribed

by the current Technical Specifications. As such, flux monitoring is

enhanced during core reloading and offloading. This change is

conservative relative to the current requirements. Therefore, no new

or different kinds of accidents are created.

3. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant reduction in a margin of safety.

VY has determined that the proposed change does not involve a

significant reduction in a margin of safety. Loading around the

geometric center of the core as prescribed by the current Technical

Specifications results in cells of moderator separating the fuel

from the instrumentation monitoring its flux. This change requires

the flux monitoring instrumentation to be in the fueled region, and,

in so doing, provides for more accurate monitoring of core flux

during core reloading and offloading. As such, the operators will

have more accurate indication and SRM automatic trip functions will

actuate when the actual flux reaches the trip setpoints. Therefore,

this change will not result in a significant reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

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

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

NRC Section Chief: James W. Clifford.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: August 18, 1999.

Description of amendment request: The licensee proposed changing

the Vermont Yankee Nuclear Power Station (VY) technical specifications

(TSs) by revising the definition of the ``Surveillance Frequency'' to

incorporate provisions that apply upon the discovery of a missed TS

surveillance. The provisions would allow 24 hours to perform the

surveillance before the applicable limiting condition for operation is

entered.

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 operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

This change does not result in any physical alteration of plant

systems, structures or components; nor does the change modify the

manner in which plant equipment will be operated or maintained. As a

result, the proposed change does not affect any of the parameters or

conditions that contribute to the initiation or mitigation of any

accidents previously evaluated.

Surveillance frequencies are not assumed in the initiation of

any analyzed event. Thus, conditions assumed in the plant accident

analyses are unchanged. Furthermore, there is no relaxation of

required setpoints or operating parameters.

Therefore, the probability or consequences of an accident

previously evaluated are not significantly increased since the most

likely outcome of performing a surveillance is that it does, in

fact, demonstrate the system or component is operable. VY has,

therefore, determined that the proposed change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

2. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not create the

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

previously evaluated. The proposed change will not modify the

physical plant or the modes of plant operation. The changes do not

involve the addition or modification of equipment nor do they alter

the design or operation of plant systems. These changes to Technical

Specifications do not create any new or different kind of accident

since they do not involve any change to the plant or the manner in

which it is operated.

Therefore, VY has determined that the proposed change does not

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

any accident previously [evaluated].

3. The operation of Vermont Yankee Nuclear Power Station in

accordance with the proposed amendment will not involve a

significant reduction in a margin of safety.

The proposed change does not affect design margins or

assumptions used in accident analyses. The capability of safety

systems to function and limiting safety system settings are

similarly unaffected as a result of this change.

The increased time allowed (up to 24 hours) for the performance

of a surveillance discovered to have not been performed, is

acceptable based on the small probability of an event requiring the

associated component. The requested allowance will provide

sufficient time to perform the missed surveillance in an orderly

manner. Without the 24 hour delay, it is possible that the missed

surveillance would force a plant shutdown; thus, the plant could be

shutting down while the missed surveillance is being performed. As a

result of this delay, the potential for human error will be reduced.

Consequently, there is no significant reduction in a margin of

safety as overall plant safety is enhanced due to the avoidance of

unnecessary plant shutdowns.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

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

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

NRC Section Chief: James W. Clifford.

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

North Anna Power Station, Unit Nos. 1 and 2, Louisa County, Virginia

Date of amendment request: August 4, 1999.

Description of amendment request: The proposed changes to North

Anna Power Station (NAPS) Units 1 and 2 Technical Specification (TS)

4.4.1.6.1 and associated Bases will extend the drained reactor coolant

loop verification time (verified as drained) from two hours to four

hours prior to backfilling when returning the drained loop to service.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated.

Administrative procedures ensure that the initiation of seal

injection in order to establish a partial vacuum in an isolated and

drained loop will not create the potential for an inadvertent and

undetected introduction of under-borated water into an isolated loop

prior to returning the isolated loop to service. Additionally,

extension of the drained loop verification time from two hours to

four hours prior to backfill operations will not significantly

diminish confidence that the isolated and drained loop will, in

fact, be drained at the time the back-fill evolution is initiated.

Therefore, there is no measurable increase in the probability or

consequences of any accident previously evaluated.

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

kind of accident from any accident previously evaluated.

[[Page 48869]]

There are no modifications to the plant as a result of the

changes. No new accident or event initiators are created by the

initiation of seal injection in order to establish a partial vacuum

in an isolated and drained loop, and by the extension of the drained

loop verification time requirement from two hours to four hours

prior to backfill operations. Therefore, the proposed changes do not

create the possibility of any accident or malfunction of a different

type previously evaluated.

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

of safety.

The proposed changes have no effect on the safety analyses

assumptions. Changes acknowledge the establishment of seal injection

for the Reactor Coolant Pump in the isolated and drained loop as a

prerequisite for the vacuum-assisted back-fill technique and extends

the drained-loop verification time from two hours to four hours

prior to backfill operations. The two hour interval was established

to ensure that the drained loop is verified to be drained at a point

in time sufficiently close to the initiation of the back-fill

evolution such that no intervening event could occur that would

render the loop no longer drained. Relaxation of the drained loop

verification time from two hours to four hours will not

significantly diminish confidence that the isolated and drained loop

will be drained at the time the back-fill evolution is initiated.

Therefore, the proposed changes do not result in a 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: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

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: April 28, 1999.

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TS) Section 3.4.A.4 and Table 4.1-

2B for Units 1 and 2. The proposed changes would reduce the minimum

volume requirement for the refueling water chemical addition tank (CAT)

to provide additional operating margin, and also correct administrative

format errors in Table 4.1-2B.

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--Does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The probability or the consequences of an accident previously

evaluated are not increased. When the revised Safety Analysis Limit

minimum CAT volume of 3800 gallons was implemented, consideration

was given to the effects of the proposed reduced CAT volume on

containment integrity analyses, containment spray and post-LOCA sump

pH analyses, and the post-LOCA recirculation switchover time

interval specified in Emergency Operating Procedures. The change was

determined to be acceptable as accident analyses assumptions would

continue to be met. The proposed TS minimum CAT volume (3930

gallons) includes an allowance for the CAT level Channel Statistical

Allowance (CSA), so that the safety analysis limit CAT volume (3800

gallons) will not be violated when the measured CAT volume (i.e.,

tank level) is at or above the TS minimum CAT volume limit. The

proposed reduction in the TS minimum CAT volume has no bearing on

the probability of occurrence of any accident previously evaluated,

since neither the volume nor the sodium hydroxide inventory of the

CAT have any bearing on postulated accident initiators. Furthermore,

because the affected accident analyses have been evaluated and found

to meet their acceptance criteria with the reduced safety analysis

limit CAT volume, the consequences of an accident previously

evaluated is not increased.

Criterion 2--Does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

The possibility of a new or different kind of accident than any

accident previously evaluated is not created. The proposed reduction

in the TS minimum CAT volume does not involve any alterations to the

physical plant that would introduce any new or unique operational

modes or accident precursors. Only the TS minimum CAT volume is

being changed to establish an operationally feasible alarm setpoint

to provide the operators additional flexibility in maintaining the

required CAT volume.

Criterion 3--Does not involve a significant reduction in a

margin of safety.

The margin of safety is not reduced. It was determined that the

affected safety analyses continue to meet their respective

acceptance criteria with the revised minimum CAT volume. By

implementing the proposed change in the TS minimum CAT volume, a CAT

level alarm setpoint may be established which includes a

conservative allowance for level measurement uncertainty such that

neither the proposed TS minimum CAT volume nor the Safety Analysis

Limit CAT volume will be violated at the time a CAT level alarm is

received. Therefore, it is concluded that the proposed change will

not reduce the margin of safety.

This analysis demonstrates that the proposed amendment to the

Surry Units 1 and 2 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.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek

Generating Station, Coffey County, Kansas

Date of amendment request: December 29, 1998, as supplemented by

letter dated July 29, 1999. The December 29, 1998, amendment

application was previously noticed in the Federal Register on February

24, 1999 (64 FR 9023).

Description of amendment request: The amendment would revise

Section 5.6.6, ``Reactor Coolant System (RCS) Pressure and Temperature

Limits Report (PTLR),'' of the improved Technical Specifications (TSs),

that were issued in Amendment 123 on March 31, 1999. The amendment

would (1) add the phrase ``and Cold Overpressure Mitigation System'' to

the first sentence of item 5.6.6.b that identifies the limits that can

be determined by the licensee in the PTLR, and (2) replace the current

list of documents listed in item 5.6.6.b by the NRC letter that will

approve this amendment and the Westinghouse report, WCAP-14040-NP-A,

``Methodology Used to Develop Cold Overpressure Mitigation System

Setpoints and RCS Heatup and Cooldown Limit Curves,'' dated January

1996. WCAP-14040-NP-A is the NRC-approved topical report that provides

a methodology for developing the cold overpressure mitigation system

(COMS) setpoints and RCS heatup and cooldown limit curves for

Westinghouse plants,

[[Page 48870]]

such as Wolf Creek Generating Station (WCGS).

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.

Incorporating the revised heatup and cooldown pressure/

temperature limit curves and the COMS PORV setpoint limit curve into

the WCGS Technical Specifications does not affect the probability or

consequences of an accident previously evaluated.

The revised limit curves are calculated using the most limiting

RTNDT for the reactor vessel components and include a

radiation-induced shift corresponding to the end of the period for

which the curves are generated. The COMS PORV Setpoint Limit Curve

is calculated using the most limiting mass injection transient,

taking into account operation of the NCP [normal charging pump]

during shutdown modes. The changes do not affect the basis,

initiating events, chronology, or availability/operability of safety

related equipment required to mitigate transients and accidents

analyzed for WCGS.

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

or different kind of accident from any accident previously

evaluated.

Adopting the revised limit curves redefines the range of

acceptable operation for the Reactor Coolant System. This

redefinition is a result of the analysis of reactor vessel

surveillance specimens removed from the reactor in a continuing

surveillance program which monitors the effects of neutron

irradiation on the WCGS reactor vessel materials under actual

operating conditions. Included in the revised limit curves is

consideration for NCP operation during shutdown modes. Incorporating

these revised curves does not create the possibility of an accident

of a different type from any previously evaluated for WCGS.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The revision of these limit curves continues to maintain the

margin of safety required for prevention of non-ductile failure of

the WCGS reactor vessel during low temperature operation as required

by 10 CFR 50, Appendices G and H. The revised curves primarily

affect RCS operation below 350 deg.F by limiting the available

pressure/temperature window for heatup and cooldown. The revised

limit curves compensate for the in-service radiation induced

embrittlement of the reactor vessel and accounts for the requirement

that the closure flange region temperature must exceed the nil-

ductility temperature by at least 120 deg.F when pressure exceeds

20% of the preservice hydrostatic test pressure.

The revised COMS PORV Setpoint Limit Curve, which includes

consideration of NCP operation during shutdown modes, ensures

overpressure protection of the RCS and reactor vessel.

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 locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621.

Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.

NRC Section Chief: Stephen Dembek.

Previously Published Notice of Consideration of Issuance of

Amendments to Facility Operating Licenses, Proposed No Significant

Hazards Consideration Determination, and Opportunity for a Hearing

The following notice was previously published as a separate

individual notice. The notice content was the same as above. It was

published as an individual notice either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. It is repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

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

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

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

and 3, York County, Pennsylvania

Date of amendment request: August 6, 1999.

Brief description of amendment request: The proposed amendments

would revise the Technical Specifications (TSs) contained in Appendix A

to the Operating Licenses to incorporate a note into the TSs which will

permit a one-time exemption, until September 30, 1999, from the

90 deg.F limit stated in Surveillance Requirement (SR) 3.7.2.2. This SR

currently requires that the average water temperature of the normal

heat sink be less than or equal to 90 deg.F as demonstrated on a 24-

hour frequency. As stated in the proposed TS note, during the time

period between approval and September 30, 1999, the average water

temperature of the normal heat sink will be limited to less than or

equal to 92 deg.F.

Date of publication of individual notice in Federal Register:

August 13, 1999 (64 FR 44243).

Expiration date of individual notice: 14 days for comments, August

27, 1999; 30 days for hearing, September 13, 1999.

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.

Notice of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

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

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

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

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

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

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

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

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

it is so indicated.

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

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

related letter, Safety Evaluation and/or Environmental Assessment as

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

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

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

the particular facilities involved.

[[Page 48871]]

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: July 30, 1999.

Brief description of amendment: The amendment revises Technical

Specification (TS) 3.7.8, ``Ultimate Heat Sink (UHS),'' to permit a 72-

hour delay in the UHS temperature restoration period prior to entering

the plant shutdown required actions. This TS amendment is given as a

temporary amendment change effective until September 30, 1999, after

which the TS will revert back to the original TS provisions.

Date of issuance: August 24, 1999.

Effective date: August 24, 1999.

Amendment No.: 184.

Facility Operating License No. DPR-23: Amendment revised the

Technical Specifications.

Public comments requested as to proposed no significant hazards

consideration (NSHC): Yes (64 FR 43406 dated August 10, 1999). The

notice provided an opportunity to submit comments on the Commission's

proposed NSHC determination. No comments have been received. The notice

also provided for an opportunity to request a hearing by September 8,

1999, but indicated that if the Commission makes a final NSHC

determination, any such hearing would take place after issuance of the

amendment.

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

exigent circumstances, and final determination of NSHC are contained in

a Safety Evaluation dated August 24, 1999.

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.

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 25, 1999.

Brief description of amendments: The amendments revise various

parts of the Technical Specifications (Appendix A of the Catawba

operating licenses) to identify that the Trip Setpoints for the reactor

trip system and engineered safety feature actuation system

instrumentation are in reality Nominal Trip Setpoints.

Date of issuance: August 13, 1999.

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

within 45 days from the date of issuance.

Amendment Nos.: 179--Unit 1; 171--Unit 2.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: May 5, 1999 (64 FR

24195).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 13, 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. 50-412, Beaver Valley Power

Station, Unit 2, Shippingport, Pennsylvania

Date of application for amendment: June 18, 1996, as supplemented

December 12, 1997, February 23, June 15, and July 15, 1999; and by

separate application dated October 22, 1997, as supplemented February

23, June 28, and July 15, 1999.

Brief description of amendment: This amendment implements: (1)

voltage-based repair criteria for BVPS-2 steam generator tubes similar

to the changes approved for BVPS-1 in License Amendment No. 198. The

changes revise BVPS-2 technical specifications (TSs) 4.4.5 and 3.4.6.2

and associated Bases to reflect the guidance provided in the Nuclear

Regulatory Commission's (NRC) Generic Letter 95-05, ``Voltage-Based

Repair Criteria for Westinghouse Steam Generator Tubes Affected by

Outside Diameter Stress Corrosion Cracking,'' (GL 95-05). Additionally,

BVPS-2 TS Table 4.4-2 is revised to reference TS 6.6 for reporting

requirements. (2) reduced reactor coolant system (RCS) specific

activity limits in accordance with the NRC's guidance provided in GL

95-05. The definition of Dose Equivalent I-131 is replaced with the

Improved Standard TS definition in the first sentence, and an equation

is added based on dose conversion derived from the International

Commission on Radiation Protection (ICRP) ICRP-30. TS 3.4.8, Specific

Activity, is revised by reducing the Dose Equivalent I-131 limit from

1.0 [micro] Ci [curies]/gram to 0.35 [micro] Ci [curies]/gram for the

48-hour limit and from 60 [micro] Ci [curies]/gram to 21 [micro] Ci

[curies]/gram for the maximum instantaneous limit. Item 4.a in TS Table

4.4-12, Primary Coolant Specific Activity Sample and Analysis Program;

TS Figure 3.4-1, and the Bases for TS 3/4.4.8 are also modified to

reflect the reduced Dose Equivalent I-131 limit.

The February 23, 1999, letter provided a revised control room dose

calculation in support of both the June 18, 1996, and October 22, 1997,

amendment requests. Importantly, this calculation assumed the lower

allowable primary-to-secondary leak rate limit associated with the June

18, 1996, submittal, and the reduced RCS specific activity limits

associated with the October 22, 1997, submittal. Because of this

interdependence, the changes of the first amendment request must be

implemented concurrently with those of the second in order for the

supporting analysis to remain valid. Hence, both of these license

amendment requests have been combined into this single amendment.

Date of issuance: August 18, 1999.

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

within 60 days.

Amendment No: 101.

Facility Operating License No. NPF-73. Amendment revised the

Technical Specifications.

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

FR 64109) and March 25, 1998 (63 FR 14485). The December 12, 1997,

February 23, June 15, June 28, and July 15, 1999, letters provided

additional information but did not change the initial proposed no

significant hazards consideration determinations or expand the

amendment requests beyond the scope of the Federal Register notices.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 18, 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: November 24, 1998, as supplemented by

letters dated February 25 and July 14, 1999.

Brief description of amendments: The amendments revise the

administrative sections of the Technical Specifications to reflect the

approved consolidated quality assurance program, clarify the

responsibilities of the shift technical advisor position on shift,

simplify the contents of the monthly operating report description,

complete the relocation of the fire protection requirements from

[[Page 48872]]

the Technical Specifications, and replace selected position titles with

descriptions of functional responsibility.

Date of issuance: August 26, 1999.

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

within 60 days from the date of issuance.

Amendment Nos.: 198 and 209.

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

revised the Technical Specifications.

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

4156).

The February 25 and July 14, 1999, letters provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 26, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

Florida Power and Light Company, Docket No. 50-335, St. Lucie Plant,

Unit No. 1, St. Lucie County, Florida

Date of application for amendment: November 22, 1998.

Brief description of amendment: This amendment revises the reactor

thermal margin safety limit lines and flow rates stated in the St.

Lucie, Unit 1, technical specifications (TS). The amendment also

updates the reference for dose conversion factors used in Dose

Equivalent Iodine-131 calculations, makes administrative changes to the

criticality analysis uncertainty described in TS 5.6.1.a.1, updates the

analytical methods used in determining core operating limits listed in

TS 6.9.1.11, and revises the TS Bases for the steam generator pressure-

low trip setpoint.

Date of Issuance: August 18, 1999.

Effective Date: August 18, 1999.

Amendment No.: 163.

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

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

FR 6696).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 18, 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.

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 December 31, 1998, and May 12, 1999.

Brief description of amendment: The amendment approves changes to

the Improved Technical Specifications to reflect the use of Topical

Report BAW-2421 for fluence determination and changes to the low

temperature over-pressure protection limits. Changes to the CR-3

Pressure/Temperature Limits Report to reflect plant operation to 32

Effective Full Power Years were included in the submittal.

Date of issuance: August 12, 1999.

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

commencing Cycle 12 operation.

Amendment No.: 183.

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

Technical Specifications.

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

FR 71965). The supplemental letters dated December 31, 1998, and May

12, 1999, 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 August 12, 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

Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: November 30, 1998.

Brief description of amendment: The Amendment revises Technical

Specifications (TS) to allow both doors of the containment personnel

air lock to be open during fuel movement and adds a provision for an

outage equipment hatch.

Date of issuance: August 16, 1999.

Effective date: August 16, 1999.

Amendment No.: 184.

Facility Operating License No. DPR-31: Amendment revised the TS.

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

4157).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 16, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal River, Florida 34428.

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

Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

Date of application for amendment: July 30, 1998, as supplemented

April 8 and July 8, 1999.

Brief description of amendment: Revises Technical Specifications

for the Control Room Emergency Ventilation System and the Ventilation

Filter Test Program.

Date of issuance: August 23, 1999.

Effective date: August 23, 1999.

Amendment No.: 185.

Facility Operating License No. DPR-31: Amendment revised the

Technical Specifications.

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

FR 64115). The April 8 and July 8, 1999, supplements did not change the

original proposed no significant hazards determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 23, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal River, Florida 34428.

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: December 3, 1998, as

supplemented by letters dated March 26, April 16, May 7, May 21, June

4, June 15, and June 29, 1999.

Brief description of amendment: The amendment revises the Technical

Specification Figure 2.1-1 ``Core Protection Safety Limit,'' and Figure

2.1-3 ``Core Protection Safety Bases'' to reflect a decrease in reactor

coolant system flow resulting from a revised analysis to allow

operation of the TMI-1 facility with an average of 20 percent of the

steam generator tubes plugged, and no more than 25 percent plugged in

either generator.

Date of issuance: August 19, 1999.

Effective date: As of the date of demonstration of a satisfactory

emergency feedwater pump flow test, as described in the license

amendment and documented by the licensee, to be

[[Page 48873]]

performed during the 13R refueling outage scheduled to begin September

10, 1999, and shall be implemented within 30 days of that date.

Amendment No.: 214.

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

Technical Specifications.

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

FR 71967). The supplements dated March 26, April 16, May 7, May 21,

June 4, June 15, and June 29, 1999, are within the scope of the

original notice and do not change the proposed no significant hazards

consideration finding.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 19, 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 1, Dauphin County, Pennsylvania

Date of application for amendment: February 2, 1999 as supplemented

July 29, 1999.

Brief description of amendment: The amendment expands the scope of

systems and test requirements for post-accident reactor building sump

recirculation engineered safeguards features systems and increases the

maximum allowable leakage of TS 4.5.4 from 0.6 gallons per hour (gph)

to 15.0 gph.

Date of issuance: August 24, 1999.

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

within 120 days.

Amendment No.: 215.

Facility Operating License No. DPR-50. This amendment revised the

Technical Specifications.

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

14283).

The supplemental letter did not change the initial no significant

hazards consideration determination or the Federal Register notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 24, 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 Nos. 50-336 and 50-

423, Millstone Nuclear Power Station, Unit Nos. 2 and 3, New London

County, Connecticut

Date of application for amendment: March 5, 1999.

Brief description of amendment: The amendments relocate certain

Technical Specifications (TSs) Section 6.0 administrative controls to

the NRC-approved Northeast Utilities Quality Assurance Program (NUQAP)

Topical Report. Specifically, Sections 6.2.3 (Unit 3 only), 6.5, 6.6

(partial), 6.7 (partial), and 6.10. The amendments also delete parts of

Section 6.6 and 6.7 because their requirements are duplicated in

existing regulations or elsewhere in the TSs. In addition, the

amendments modify the table of contents and other TS sections to

incorporate the aforementioned changes (e.g., correct references).

Date of issuance: August 13, 1999.

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

within 60 days from the date of issuance.

Amendment Nos.: 239 and 173.

Facility Operating License Nos. DPR-65 and NPF-49: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: April 7, 1999 (64 FR

17027).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 13, 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.

PECO Energy Company, Public Service Electric and Gas Company Delmarva

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

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

and 3, York County, Pennsylvania

Date of application for amendments: February 12, 1999, as

supplemented July 8, 1999. The July 8, 1999, letter provided clarifying

information and did not change the original no significant hazards

consideration determination.

Brief description of amendments: Administrative changes to correct

typographical and editorial errors in Technical Specifications

introduced in previous amendments.

Date of issuance: August 23, 1999.

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

of issuance. The amendment will be implemented within 30 days.

Amendments Nos.: 228 and 231.

Date of initial notice in Federal Register: May 5, 1999 (64 FR

24200).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 23, 1999.

No significant hazards consideration comments received: No.

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.

PP&L, Inc., Docket Nos. 50-387 and 50-388, Susquehanna Steam Electric

Station, Units 1 and 2, Luzerne County, Pennsylvania

Date of application for amendments: November 20, 1998, as

supplemented by letter dated June 25, 1998.

Brief description of amendments: These amendments modified

technical specification surveillance requirement, 3.8.1.4, to allow

increases in the minimum fuel oil required to be stored in the day

tanks for emergency diesel generators.

Date of issuance: August 23, 1999.

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

within 30 days.

Amendment Nos.: 185 and 159.

Facility Operating License Nos. NPF-14 and NPF-22. The amendments

revised the Technical Specifications.

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

4159).

The supplemental letter provided clarifying information and did not

change the initial no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 23, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Osterhout Free Library,

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

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

362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego

County, California

Date of application for amendments: December 19, 1997, as

supplemented June 1, 1998, and May 13, 1999.

[[Page 48874]]

Brief description of amendments: The amendments revise TS 3.4.9,

Pressurizer, to reduce the allowable pressurizer water volume for

pressurizer operability. The allowable water volume is also revised to

a percent pressurizer level of 57 percent.

Date of issuance: August 19, 1999.

Effective date: August 19, 1999, to be implemented within 30 days

of issuance.

Amendment Nos.: Unit 2--155; Unit 3--146.

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

revised the Technical Specifications.

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

14488).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 19, 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.

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

362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego

County, California

Date of application for amendments: September 4, 1998, as

supplemented December 8, 1998, and February 16, 1999 (PCN 493).

Brief description of amendments: The amendments revise Technical

Specification 3.4.10, Pressurizer Safety Valves, to increase the as-

found pressurizer safety valve setpoint tolerances.

Date of issuance: August 19, 1999.

Effective date: August 19, 1999, to be implemented within 30 days

of issuance.

Amendment Nos.: Unit 2--156; Unit 3--147.

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

revised the Technical Specifications.

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

FR 6711). The licensee's letters dated December 8, 1998, and February

16, 1999, provided clarifications and additional information that were

within the scope of the original Federal Register notice and did not

change the staff's initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 19, 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.

Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph M.

Farley Nuclear Plant, Unit 1, Houston County, Alabama.

Date of amendment request: April 23, 1999, as supplemented by

letters dated July 22, July 30 and August 12, 1999.

Brief Description of amendment: The amendment adds an additional

condition to the license which allows Southern Nuclear Operating

Company to operate Unit 1 for Cycle 16 based on a risk-informed

approach to evaluate steam generator tube structural integrity.

Date of issuance: August 17, 1999.

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

within 30 days from the date of issuance.

Amendment No.: 143.

Facility Operating License No. NPF-2: Amendment revises the

Facility Operating License to add a license condition.

Date of initial notice in Federal Register: June 16, 1999 (64 FR

32291).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 17, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.

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 22, 1999, as supplemented July 15,

1999.

Brief description of amendments: The amendments revised Technical

Specification 3/4.7.1.6, ``Atmospheric Steam Relief Valves,'' and added

a new Technical Specification for atmospheric steam relief valve

instrumentation, to ensure that the automatic feature of the steam

generator power-operated relief valves (i.e., the atmospheric steam

relief valves) remains operable during Modes 1 and 2.

Date of issuance: August 19, 1999.

Effective date: August 19, 1999, to be implemented within 30 days.

Amendment Nos.: Unit 1--114; Unit 2--102.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: April 21, 1999 (64 FR

19565).

The July 15, 1999, supplement provided revised Technical

Specification pages and clarifying information that was within the

scope of the original Federal Register notice and did not change the

staff's initial proposed no significant hazards consideration

determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated August 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 Nos. 50-260 and 50-296, Browns Ferry

Nuclear Plant, Units 2 and 3, Limestone County, Alabama

Date of application for amendment: June 3, 1999 (TS 397).

Brief description of amendment: The Amendments change the Technical

Specifications (TS) by reducing the Allowable Value used for Reactor

Vessel Water Level--Low, Level 3 for several instrument functions.

Date of issuance: August 16, 1999.

Effective date: August 16, 1999.

Amendment Nos.: 260 and 219.

Facility Operating License Nos. DPR-52 and DPR-68: Amendments

revise the TS.

Date of initial notice in Federal Register: July 14, 1999 (64 FR

38037).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated August 16, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Athens Public Library, 405 E.

South Street, Athens, Alabama 35611.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: April 16, 1999, as supplemented

June 9, 1999.

Brief description of amendment: The amendment clarifies the

inservice inspection requirements regarding the granting of relief from

the American Society of Mechanical Engineers (ASME) Code requirements

by the NRC. The amendment also made changes to reflect previous NRC

approval of the use of ASME Code Case N-560.

Date of Issuance: August 13, 1999.

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

implemented within 30 days.

[[Page 48875]]

Amendment No.: 172.

Facility Operating License No. DPR-28. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 14, 1999 (64 FR

38037).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated August 13, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

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

Yankee Nuclear Power Station, Vernon, Vermont

Date of application for amendment: June 24, 1999.

Brief description of amendment: The amendment clarifies the basis

for the reactor protection system bypass of the turbine stop valve

closure and turbine control valve fast closure scram signals at low

power. The amendment clarifies that the analytical basis for this

bypass corresponds to a fraction of reactor rated thermal power and not

other measures of power, for instance, turbine power.

Date of Issuance: August 13, 1999.

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

implemented within 30 days.

Amendment No.: 173.

Facility Operating License No. DPR-28.: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 14, 1999 (64 FR

38038).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated August 13, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Dated at Rockville, Maryland, this 1st day of September 1999.

For the Nuclear Regulatory Commission.

Suzanne C. Black,

Deputy Director, Division of Licensing Project Management, Office of

Nuclear Reactor Regulation.

[FR Doc. 99-23300 Filed 9-7-99; 8:45 am]

BILLING CODE 7590-01-P