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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from May 8, 1999, through May 20, 1999. The last

biweekly notice was published on May 19, 1999.

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of Administration Services, Office of

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

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

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

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

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

written comments received may be examined at the NRC Public Document

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

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

is discussed below.

By July 2, 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.

[[Page 29708]]

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.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power

Station, Plymouth County, Massachusetts

Date of amendment request: May 5, 1999.

Description of amendment request: The proposed amendment would

change the technical specifications (TS) and licensing basis for the

required amount of diesel fuel to be stored on-site and its sources.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change affects only the on-site diesel fuel storage

capacity for the operation of emergency diesel generators [EDG]. The

on-site storage capacity is not associated with an accident

precursor/initiator; thus, it has no impact on the probability of

[an] accident occurring. The consequences of an accident would not

be significantly increased because reasonable measures will be

available to ensure the EDGs are supplied with enough fuel from the

on-site sources to operate for seven days at rated capacity.

The proposed amendment does not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The proposed change does not affect normal plant operation or

the immediate response to an accident. The only change is the

proposed refilling operation to transfer fuel from the Class II

SBODG [Station Blackout Diesel Generators] storage tanks to the

Class I EDG tanks. The refilling operation would occur entirely

outdoors through above ground hoses connecting the EDG and SBODG

tanks. This operation would only be required following a LOCA [loss-

of-coolant accident], an accident already analyzed. Since the

proposed refilling operation is a post-accident evolution, it would

not be in place to cause an accident of a different type during non-

accident conditions. No reasonable malfunction of equipment

associated with the evolution could create a new or different kind

of accident than previously evaluated.

The proposed amendment does not involve a significant reduction

in the margin of safety.

The proposed amendment for licensing basis change and TS change

does not significantly reduce the margin of safety. The proposed

change restores the licensing basis to provide sufficient fuel in

on-site storage tanks for continuous operation of each EDG for

approximately seven days. The revised licensing basis requires

36,800 gallons of fuel per EDG to be stored on-site. A minimum of

19,800 gallons of fuel will be stored in Class I EDG storage tanks

and the remaining will be stored in Class II SBODG on-site storage

tanks. The storage of fuel in Class I tanks does not reduce the

margin of safety. The only potential reduction in the margin of

safety is due to the use of Class II SBODG tanks and associated

transfer equipment for the storage and transfer of additional fuel.

These Class II tanks are rugged, double-wall fiberglass tanks. While

not designed to safety-related requirements, the failure of these

tanks under extreme environmental conditions, such as an earthquake,

has been evaluated to be very unlikely. Thus, on-site storage of

sufficient fuel for operation of both EDGs is assured to mitigate

the consequences of an accident previously evaluated. All stored

fuel is maintained at the same quality standard. The proposed diesel

fuel refilling operation is a post design basis accident activity,

which does not create the possibility of a new accident or impact an

accident previously evaluated. Therefore, there is no significant

reduction in the safety margin.

The NRC 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: Plymouth Public Library, 132

South Street, Plymouth, Massachusetts 02360.

Attorney for licensee: J. Fulton, Boston Edison Company, 800

Boylston Street, 36th Floor, Boston, Massachusetts 02199.

NRC Section Chief: James W. Clifford.

[[Page 29709]]

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station, Units 1 and 2, Lake County, Illinois

Date of application of amendment request: October 2, 1998, as

supplemented by letter dated April 19, 1999.

Description of amendment request: By letter dated February 13,

1998, Commonwealth Edison Company (ComEd) certified that they have

permanently ceased operations at Zion Nuclear Power Station (ZNPS),

Units 1 and 2. Since ComEd has permanently ceased operations at ZNPS,

they have requested an amendment to the Facility Operating Licenses to

eliminate license conditions that are no longer applicable and to

replace the existing technical specifications in their entirety with

permanently defueled technical specifications (PDTS). The PDTS reflect

the permanently shutdown and defueled condition of the ZNPS.

Basis for a 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 and has determined that the proposed changes do not:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The administrative changes remove requirements that are not

invoked with the reactors permanently defueled. The editorial

changes alter format, word choice, grammar, terminology, etc., but

do not change requirements. The more restrictive changes add new

requirements, remove existing exceptions, or make existing limits

more conservative. The relocation or redundancy changes remove

requirements from the facility operating licenses or technical

specifications because they exist in another document controlled by

other approved methods. None of these types of changes affect the

probability or consequences of a previously evaluated accident since

there is no functional reduction in the limitations imposed on

structures, systems, components or activities with the reactors

permanently defueled.

The less restrictive changes to the license conditions eliminate

requirements for programs and commitments that address hazards or

conditions that are no longer credible with both reactors

permanently defueled. Since these hazards or conditions are not

credible, no increase in the probability or consequences of a

previously evaluated accident will result from the elimination of

these requirements.

The less restrictive changes to the equipment-related technical

specifications eliminate or modify restrictions involving certain

structures systems and components (SSCs). Some of the equipment-

related technical specifications have been eliminated because, with

both reactors permanently defueled, the spectrum of previously

evaluated credible accidents has been significantly reduced and many

of the associated hazards (such as reactor coolant gaseous activity,

hydrogen, and radioactive iodine) will not occur. Since those

previously evaluated accidents and associated hazards are no longer

credible, their probability and consequences are not increased by

the changes eliminating the associated technical specifications.

Other equipment-related technical specifications have been modified

to address previously evaluated accidents that are still relevant in

the permanently defueled condition more logically and consistently,

without increasing their probability or consequences.

The less restrictive changes to the Administrative Control

technical specifications affect a variety of functions. They provide

flexibility in Quality Assurance Program administration, allow a

reduction in shift staffing, eliminate certain training requirements

for personnel who have little or no safety involvement, change

certain procedure processing requirements, provide consistency in

scheduling certain radiological surveillances and reports, eliminate

reports that are no longer needed, eliminate unnecessary flood door

requirements, and allow alternative methods of administering Process

Control Program changes. Since none of these changes directly

involve the previously evaluated accidents that remain credible with

both reactors permanently defueled, the changes will not increase

the probability or consequences of any previously evaluated

accident.

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

or different kind of accident from any accident previously

evaluated. The administrative changes do not alter any SSCs or

activities involved with the safe storage of nuclear fuel. The

editorial changes do not alter any requirements. The more

restrictive changes make the technical specifications more limiting.

The relocation/redundancy changes only change the location of

requirements. None of these types of changes create the possibility

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

evaluated.

The less restrictive changes to the license conditions eliminate

requirements for programs and commitments involving hazards or

conditions that are no longer credible with both reactors

permanently defueled. Since these changes do not result in any new

programs or activities, they do not create the possibility of a new

or different kind of accident from any accident previously

evaluated.

The less restrictive changes to the equipment-related technical

specifications do not alter any SSC or cause any SSC to be operated

in a manner that could initiate any event or accident. Therefore,

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

kind of accident from any accident previously evaluated.

The less restrictive changes to the Administrative Control

technical specifications do not change the design, function, or

operation of any SSC except the flood doors and the change involving

the flood doors does not introduce any new type of event. Therefore,

the less restrictive changes to the Administrative Control technical

specifications do not create the possibility of a new or different

kind of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The administrative changes do not alter any SSCs or activities

involved with the safe storage of nuclear fuel. The editorial

changes do not alter any requirements. The more restrictive changes

make the technical specifications more limiting. The relocation/

redundancy changes only change the location of requirements. None of

these types of changes reduce any safety margin.

The less restrictive changes to the license conditions eliminate

requirements that apply to hazards or conditions that are no longer

relevant with both reactors permanently defueled. The safety margins

that may have been associated with those license conditions are no

longer relevant.

There are no longer any relevant margins of safety associated

with the less restrictive changes to the equipment-related technical

specifications except for those involving criticality control and

seismic criteria. The proposed technical specifications maintain the

same margin of safety for criticality control in the spent fuel

pool, and the Defueled Safety Analysis Report imposes seismic

criteria that provide an adequate safety margin.

The less restrictive changes to the Administrative Control

technical specifications do not directly involve any limits or

parameters and therefore cannot affect any 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: Waukegan Public Library, 128

N. County Street, Waukegan, Illinois 60085.

Attorney for licensee: Pamela B. Strobel, Senior Vice President and

General Counsel, Commonwealth Edison Company, P.O. Box 767, Chicago,

Illinois 60690-076.

NRC Project Director: Stuart A. Richards.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit

No. 1, Pope County, Arkansas

Date of amendment request: May 14, 1999.

Description of amendment request: The proposed amendment would

revise the Technical Specification requirements affecting the

surveillance criteria for that portion of the once-through steam

generator tubes regarded

[[Page 29710]]

as a primary-to-secondary pressure boundary located within the upper

tube sheet and impacted by a specific degradation mechanism, namely,

outside diameter intergranular attack.

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 once-through steam generators OTSG are used to remove heat

from the reactor coolant system during normal operation and during

accident conditions. The OTSG tubing forms a substantial portion of

the reactor coolant pressure boundary. An OTSG tube failure is a

breach of the reactor coolant pressure boundary and is a specific

accident analyzed in the ANO-1 [Arkansas Nuclear One, Unit 1] Safety

Analysis Report.

The purpose of the periodic surveillance performed on the OTSGs

in accordance with ANO-1 Technical Specification (TS) 4.18 is to

ensure that the structural integrity of this portion of the reactor

coolant system will be maintained. The TS plugging limit of 40% of

the nominal tube wall thickness requires tubes to be repaired or

removed from service because the tube may become unserviceable prior

to the next inspection. Unserviceable is defined in the TS as the

condition of a tube if it leaks or contains a defect large enough to

affect its structural integrity in the event of an operating basis

earthquake, a loss-of-coolant accident, or a steam line or feedwater

line break. The proposed TS change allows OTSG tubes with ODIGA

[outside diameter intergranular attack] indications contained within

a defined area of the UTS [upper tube sheet] to remain in service

with existing degradation exceeding the existing 40% through-wall

(TW) plugging limit.

Extensive testing and plant experience has illustrated that

ODIGA flaws confined to this area within the OTSG will not result in

tube burst or tube leakage. Therefore, allowing ODIGA flaws in this

specific region to remain in service will not alter the conditions

assumed in the current ANO-1 accident analysis for OTSG tube

failures under postulated accident conditions. In addition, the

condition of the OTSG tubes in this region are monitored during

regular inspection intervals to assess for evidence of growth. Any

growth noted will be addressed through testing and the operational

assessment * * *.

Application of the ODIGA alternate repair criteria will allow

leaving tubes with ODIGA indications found in the defined area of

the UTS in service while ensuring safe operation by monitoring and

assessing the present and future conditions of the tubes. ANO-1 has

operated since 1984 with ODIGA affected tubes in service with no

appreciable effect on structural integrity or indications of tube

leakage from ODIGA sources within the UTS. Through the inspection,

testing, monitoring, and assessment program previously mentioned,

and the on-line leak detection capabilities available during plant

operation, continued safe operation of ANO-1 is reasonably assured.

Therefore, the application of the ODIGA alternate repair

criteria...does not involve a significant increase in the

probability or consequences of any accident previously evaluated.

Criterion 2--Does Not Create the Possibility of a New or Different Kind

of Accident from any Previously Evaluated

The implementation of the ODIGA alternate repair criteria will

not result in any failure mode not previously analyzed. The OTSGs

are passive components. The intent of the TS surveillance

requirements are being met by these proposed changes in that

adequate structural and leak integrity will be maintained.

Additionally, the proposed change does not introduce any new modes

of plant operation.

Therefore, this change does not create the possibility of a new

or different kind of accident from any previously evaluated.

Criterion 3--Does Not Involve a Significant Reduction in the Margin of

Safety

The application of an alternate repair criteria for ODIGA

provides adequate assurance with margin that ANO-1 steam generator

tubes will retain their integrity under normal and accident

conditions. The structural requirements of ODIGA affected tubes have

been evaluated satisfactorily and meet or exceed regulatory

requirements. Leakage rates for these tubes within the defined

region of the upper tubesheet are essentially zero and are

reasonably assured to remain within the assumptions of the accident

analysis by proper application of the ODIGA alternate repair

criteria program. Because no appreciable impact is evidenced on the

tubes structural integrity or its resulting leak rate, the margin to

safety remains effectively unaltered.

Therefore, this change does not involve a significant reduction

in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

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

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

NRC Section Chief: Robert A. Gramm.

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

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

Date of amendment request: May 5, 1999.

Description of amendment request: The proposed amendment would

revise the Crystal River Unit 3 (CR-3) Improved Technical

Specifications to approve an alternate repair criteria (ARC) for axial

tube end crack-like indications in the upper and lower tubesheets of

the CR-3 Once Through Steam Generators (OTSGs). The ARC will allow

leaving OTSG tubes with axially oriented tube end cracks located within

the clad region of the tube-to-tubesheet roll joint in service. Tubes

with crack-like indications within the carbon steel portion of the

tubesheet, or tubes with circumferentially oriented tube end cracks or

volumetric indications within the Inconel clad region of the tubesheet,

would be repaired or removed from 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. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

This LAR [License Amendment Request] proposes to implement an

alternate repair criteria (ARC) for Once Through Steam Generator

(OTSG) tubes with axial tube end crack (TEC) indications.

Application of the ARC will allow tubes with axially oriented TEC to

remain in service in accordance with specific conditions. Based on a

combination of structural analyses, mock-up testing and inservice

inspections, as detailed in Topical Report BAW-2346P, allowing tubes

with TEC indications to remain in service is safe and justified.

Potential leakage from tubes with TEC will be bounded by the

main steam line break (MSLB) evaluation presented in the Final

Safety Analysis Report (FSAR). The proposed change requires

inspections during subsequent outages of tubes remaining in-service

with the TEC indications. The addition of this inspection does not

change any accident initiators. The proposed inspection of these

indications during the subsequent OTSG inservice inspections assures

continuous monitoring of these tubes such that degradation of tubes

containing TEC indications will be detected. Therefore, this change

does not involve a significant increase in the probability or

consequences of any accident previously evaluated.

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

from any accident previously evaluated.

The proposed alternate repair criteria for axial TEC indications

introduces no new failure modes or accident scenarios. Topical

Report BAW-2346P demonstrated structural and leakage integrity for

all normal operating and accident conditions for Crystal River Unit

3 (CR-3). Furthermore, leaving TEC in service does not change the

design or operating characteristics of the OTSGs. In the unlikely

event that a tube with a TEC should

[[Page 29711]]

fail and sever completely, the tube would remain engaged in the

tubesheet bore, preventing interaction with other surrounding tubes.

In this case, leakage is bounded by the steam generator tube rupture

(SGTR) accident analysis. Therefore, this change does not create a

possibility of a new or different kind of accident from any

previously evaluated.

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

The mechanical joint is constrained within the tubesheet bore;

thus, there is no additional risk associated with tube rupture. ITS

[Improved Technical Specifications] Bases 3.4.12 contains relevant

information pertaining to limitations on Reactor Coolant System

leakage. The accident leakage is shown to be less than one gallon

per minute primary-to-secondary leakage. Therefore, the FSAR

analyzed accident scenarios remain bounding, and the use of the

proposed alternate repair criteria does not reduce 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 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: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428.

Attorney for licensee: R. Alexander Glenn, General Counsel, Florida

Power Corporation, MAC-A5A, P.O. Box 14042, St. Petersburg, Florida

33733-4042.

NRC Section Chief: Sheri R. Peterson.

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

Point Plant Units 3 and 4, Dade County, Florida

Date of amendment request: April 26, 1999.

Description of amendment request: The proposed amendments would

revise the Turkey Point Plant, Units 3 and 4, Facility Operating

Licenses and the Technical Specifications (TS): (1) To remove a part of

license condition 3.L that is obsolete, (2) to update the TS Index to

reflect all changes made to the TS Sections, TS Figures, and TS Tables

by previously approved license amendments, and (3) to remove Table and

Figure numeration inconsistencies found in TS 3/4.1.2.5 and TS 3/4.7.6.

These proposed changes represent an administrative update to the Turkey

Point Plant, Units 3 and 4, Facility Operating Licenses and to the TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

amendments would not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed amendments do not involve a significant increase in

the probability or consequences of an accident previously evaluated

because the proposed changes are administrative in nature removing

obsolete references in the license conditions, updating the

Technical Specification (TS) Index to reflect the revisions made to

the TS Sections, Tables, and Figures via previous TS amendments.

These amendments will not involve a significant increase in the

probability or consequences of an accident previously evaluated

because they do not affect assumptions contained in plant safety

analyses, the physical design and/or operation of the plant, nor do

they affect Technical Specifications that preserve safety analysis

assumptions. Therefore, the proposed changes do not affect the

probability or consequences of accidents previously analyzed.

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

amendments would not create the possibility of a new or different

kind of accident from any accident previously evaluated.

The use of the modified specifications can not create the

possibility of a new or different kind of accident from any

previously evaluated since the proposed amendments will not change

the physical plant or the modes of plant operation defined in the

facility operating license. No new failure mode is introduced due to

the administrative changes since the proposed changes do not involve

the addition or modification of equipment nor do they alter the

design or operation of affected plant systems, structures, or

components.

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

amendments would not involve a significant reduction in a margin of

safety.

The operating limits and functional capabilities of the affected

systems, structures, and components are unchanged by the proposed

amendments. The proposed changes to the Facility Operating License

Conditions and to the Technical Specifications are administrative

and do not significantly reduce any of the margins 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: Florida International

University, University Park, Miami, Florida 33199.

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

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

NRC Project Director: Herbert N. Berkow.

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

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

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

and 3, York County, Pennsylvania

Date of application for amendments: March 1, 1999.

Description of amendment request: Changes are proposed to support a

modification which will install a digital Power Range Neutron

Monitoring (PRNM) system and incorporate long-term thermal-hydraulic

stability solution hardware.

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:

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

the probability or consequences of an accident previously evaluated.

As discussed in the LTR [licensing topical report], the [Nuclear

Measurements Analysis and Control] NUMAC PRNM modification and

associated changes to the TS [technical specifications] involve

equipment that is designed to detect the symptoms of certain events

or accidents and initiate mitigating actions. The worst case failure

of the equipment involved in the modification is a failure to

initiate mitigating action (scram or rod block), but no failure can

cause an accident. The PRNM replacement system is designed to

perform the same operations as the existing Power Range Monitor

system and meets or exceeds all operational requirements. Therefore,

it is concluded that the probability of an accident previously

evaluated is not increased as a result of replacing the existing

equipment with the PRNM equipment.

The PRNM system reduces the need for tedious operator actions

during normal conditions and allows the operator to focus more on

overall plant conditions. The automatic self-test and increased

operator information provided with the replacement system are likely

to reduce the burden during off-normal conditions as well. The

replacement equipment qualifications fully envelope the

environmental conditions, including electromagnetic interference, in

the PBAPS [Peach Bottom Atomic Power Station] control room.

The replacement equipment has been specifically designed to

assure that it fully meets the response time requirements in the

worst case. As a result, due to statistical variations resulting

from the sampling and update cycles, the response time is typically

faster than required in order to assure that the required response

time is always met.

[[Page 29712]]

Setpoints are changed only when justified by the improved equipment

performance specifications and by setpoint calculations which show

that safety margins are maintained. There is no impact to the

Control Rod Drop accident analysis because the PRNM system maintains

all existing system functions with a reliability equal to or better

than the existing Power Range Monitor system.

The replacement equipment includes up to 5 LPRM [Local Power

Range Monitor] inputs on a single module compared to one per module

on the current system. Up to 17 LPRM signals are processed through

one preprocessor. The recirculation flow signals are processed in

the same hardware as the LPRM processing. The net effect of these

architectural aspects is that there are some single failures that

can cause a greater loss of ``sub-functionality'' than in the

current system. Other architectural and functional aspects, however,

have an offsetting effect. Redundant power supplies are used so that

a single failure of AC power has no effect on the overall PRNM

system functions while still resulting in a half scram as does the

current system. Continuous automatic self-test also assures that if

a single failure does occur, it is much more likely to be detected

immediately. The net effect is that from a total system level,

unavailability of the safety-related functions in the replacement

system is equal to or better than the current Power Range Monitor

system.

Based on the extensive and through [sic] [thorough] verification

and validation program used in the PRNM design and field operating

experience, common cause failures in software controlled functions

are judged to not be a significant failure mode. However, in spite

of that conclusion, means are provided within the system to mitigate

the effects of such a failure and alert the operator. Therefore,

such a failure, even if it occurred, will not increase the

consequences of a previously evaluated accident.

To reduce the likelihood of common cause failure of software

controlled functions, thorough and careful verification and

validation (V&V) activities are performed both for the requirements

and the implementing software design. In addition, the software is

designed to limit the loading that external systems or equipment can

place on the system, thus significantly reducing the risk that some

abnormal dynamic condition external to the system can cause system

functional performance problems due to processing ``overload''

(i.e., ``slowing down'' or stopping the processing).

As a conservatism, however, despite these V&V activities, common

cause failures of software controlled functions due to residual

software design faults are assumed to occur. Both the software and

hardware are designed to manage the consequences of such failure

(and also cover potential common cause hardware failures). Safety

outputs are designed to be fail safe by requiring dynamic update of

output modules or data signals, where failure to update the

information is detected by simple receiving hardware, which, in

turn, forces a trip. This aspect covers all but rather complex

failures where the software or hardware executes a portion of the

overall logic but fails to process some portion of new information

(inputs ``freeze'') or some portion of the logic (outputs

``freeze'').

To help reduce the likelihood of complex failures, a watchdog

timer is used which is updated by a very simple software routine

that in turn monitors the operational cycle time of all tasks in the

system. The software design is such that as long as all tasks are

updated at the design rate, it is likely that software controlled

functions are executing as intended. Conversely, if any task fails

to update at the design rate, that is a strong indication of at

least some unanticipated condition. If such a condition occurs, the

watchdog timer will not be updated, the computer will be

automatically restarted, and the system will detect an abnormal

condition and provide an alarm and trip.

The information available to the operator is at least the same

as with the current system and, in many cases, improved. No actions

are required by the operator to obtain information normally used and

equivalent to that available with the current equipment. However,

the replacement system does provide more directly accessible

information regarding the condition of the equipment, including

automatic self-test, which can aid the operator in diagnosing

unusual situations beyond those defined in the licensing basis.

In summary, the reliability of the new PRNM system and its

ability to detect and mitigate abnormal flux transients have either

remained the same or improved over the existing Power Range Monitor

system. Since these postulated reactivity transients are mitigated

by the new system as effectively and reliability [sic] [reliably] as

the existing system, the consequences of these transients have not

changed. Therefore, the proposed changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

PBAPS Modification P00507 uses digital processing with software

(firmware) control for the main signal processing part of the

modification. The remainder of the equipment in the modification

uses conventional equipment similar to the current system (e.g.,

penetrations, cables, interface panels).

The digital equipment has ``control'' processing points and

software controlled digital processing where the current system has

analog and discrete component processing. The result is that the

specific failures of hardware and potential software common cause

failure are different from the current system. The effects of

software common cause failure are mitigated by hardware design and

system architecture, but are of a ``different type'' of failure than

those evaluated in the PBAPS Updated Final Safety Analysis Report

(UFSAR). In general, the PBAPS UFSAR assumes simplistic failure

modes (relays for example) but does not specifically evaluate such

effects as self-test detection and automatic trip or alarm.

Therefore, the replacement system may have a malfunction of a

different type from those evaluated in the PBAPS UFSAR [* * *].

However, when these PRNM failures are evaluated at the system level,

there are no new effects.

PBAPS Modification P00507 involves equipment that is intended to

detect the symptoms of certain transients and accidents and initiate

mitigating action. The worst case failure of the equipment involved

in the modification is a failure to initiate mitigating action

(scram), but no failure can cause an accident. This is unchanged

from the current system. Software common cause failures could cause

the system to fail to perform its safety function, but this

possibility is addressed in Section (i) above. In that case, it

might fail to initiate action to mitigate the consequences of an

accident, but would not cause one. No new system level failure modes

are created with the PRNM system.

Therefore, PBAPS Modification P00507 does not create the

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

previously evaluated.

iii. The proposed changes do not involve a significant reduction

in a margin of safety.

The PRNM system response time and operator information is either

maintained or improved over the current Power Range Monitor system.

The PRNM system has improved channel trip accuracy compared to

the current system and meets or exceeds system requirements assumed

in setpoint analysis. The channel response time exceeds the

requirements.

The channel indicated accuracy is improved over the current

system and meets or exceeds all of the system requirements.

The PRNM system was developed to detect the presence of thermal-

hydraulic instabilities and automatically initiate the necessary

actions to suppress the oscillations prior to violating the MCPR

Safety Limit. The NRC has reviewed and approved the LTR concluding

that the PRNM system will provide the intended protection.

Therefore, PBAPS Modification P00507 does 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: Government Publications

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

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

PA 17105.

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

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

PA 19101.

NRC Section Chief: James W. Clifford.

[[Page 29713]]

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

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

Date of amendment request: January 28, 1999, as supplemented April

29, 1999, and May 17, 1999. This notice supersedes a previous notice

(64 FR 19563) published April 21, 1999, which was based upon the

licensee's application for amendment dated January 28, 1999.

Description of amendment request: This application for amendment to

the Indian Point 3 Technical Specifications (TSs) proposes to reduce

the number of Emergency Diesel Generators (EDGs) required to be

operable during cold shutdown from 2 to 1 under certain conditions.

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 license amendment involve a significant

increase in the probability or consequences of an accident

previously analyzed?

Response: No. The equipment, which is affected by the proposed

Technical Specification change, is not an initiator to those

accidents postulated to occur during Cold Shutdown or Refueling

operating conditions. A comprehensive systems review and EDG loading

electrical analysis has demonstrated the ability of those shutdown

support systems, necessary to provide safe shutdown needs, to

perform their safety functions for the postulated accidents during

Cold Shutdown and Refueling conditions. One EDG can support the

necessary electrical loads required in Cold Shutdown and Refueling

in the event of postulated accidents along with a LOOP [loss of

offsite power] in the time frame required to prevent reactor core/

cavity/SFP [spent fuel pit] heatup concerns. This EDG support relies

upon existing plant designed manual closure of 480VAC EDS

[electrical distribution system] bus tie breakers to allow a single

EDG to pick up other 480VAC EDS bus loads, such as supplying an RHR

[residual heat removal] pump and SFP cooling pump, located on 480VAC

EDS buses 3A, 5A, or 6A. Together, operability of the required

offsite circuit(s) and one EDG along with necessary portions of the

AC, DC and 120 VAC vital instrument bus electrical power

distribution subsystems ensures the availability of sufficient

electrical sources to operate the unit in a safe manner and to

mitigate the consequences of postulated accidents during shutdown

(e.g., Fuel Handling Accidents), as well as other postulated events.

Action statements provide prompt, specific guidance to ensure

sufficiently conservative plant response should the expected EDG

power supply or required offsite power supply feeders or necessary

portions of AC, DC and 120 VAC vital instrument bus electrical power

distribution subsystems not be available. These Action Statements

are similar to those in the STS [Standard Technical Specifications].

Therefore, the proposed license amendment (i.e., changes to 3.7.F.4

and the added sections of 3.7.F.5 & 3.7.F.6) does not involve a

significant increase in the probability or consequences of an

accident previously analyzed.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response: No. The proposed license amendment does not involve

any physical changes to plant systems or component set points. The

use of 480VAC EDS bus tie breakers to power loads from necessary

energized 480VAC bus(es) is part of present plant design and

included within the present LOOP Off-Normal operating procedures

when the reactor is in Cold Shutdown operating conditions. As

discussed in the Standard Technical Specifications, NUREG 1431,

during plant shutdown with one EDG, it is not required to assume a

single failure and concurrent loss of all offsite or all onsite

power. Worst case bounding events are deemed not credible in Cold

Shutdown and Refueling conditions because the energy contained

within the reactor pressure boundary, reactor coolant temperature

and pressure, and the corresponding stresses result in the

probabilities of occurrence being significantly reduced or

eliminated, and ultimately result in minimal consequences. The lone

EDG is capable of accepting and starting required loads within the

assumed loading sequence intervals and in the time frame required to

prevent reactor core/cavity/SFP heatup concerns, with sufficient

``kW loading''. Action statements provide prompt, specific guidance

to ensure sufficiently conservative plant response should the

expected EDG or offsite supply feeder or the necessary portions of

the AC, DC and 120 VAC vital instrument bus electrical power

distribution subsystems not be available. These action statements

are similar to those in the STS. Therefore, the proposed license

amendment (i.e., changes to 3.7.F.4 and added sections 3.7.F.5 &

3.7.F.6) does not create the possibility of a new or different kind

of accident from any accident previously evaluated.

(3) Does the proposed license amendment involve a significant

reduction in the margin of safety?

Response: No. The electrical power system specifications support

the equipment required to be operable, commensurate with the current

level of safety, including the equipment requiring an EDG backed

power source. The design review results demonstrate that operation

in the conditions of Cold Shutdown and Refueling, in accordance with

the proposed Technical Specification change, is acceptable from an

accident mitigation standpoint. The basic system functions in Cold

Shutdown and Refueling operating conditions are not changed. One

EDG, along with the necessary portions of the AC, DC and 120 VAC

vital instrument electrical power distribution subsystems available,

can supply the necessary electrical power requirements during these

plant operating conditions, and in the time frame required to

prevent reactor core/cavity/SFP heatup concerns, with sufficient

``kW loading''. The analysis conducted shows that the systems are

capable of performing their design basis functions. Applicable

safety analysis in the Standard Technical Specifications, NUREG

1431, discusses these system requirements as well (i.e., it is not

required to assume a single failure and concurrent loss of all

offsite or all onsite power). Action statements, similar to those in

the Standard Technical Specifications, provide prompt, specific

operator actions to ensure sufficiently conservative plant response

should the expected EDG power supply or the required offsite power

supply feeders or AC, DC and 120 VAC vital instrument bus electrical

power distribution subsystems not be available. On this basis, the

proposed license amendment (i.e., changes to 3.7.F.4 and added

sections 3.7.F.5 & 3.7.F.6) does not involve a significant reduction

in the margin of safety.

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

this review, it appears that the three standards of 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 10601.

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

York, New York 10019.

NRC Section Chief: S. Singh Bajwa.

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

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

Date of amendment request: April 6, 1999.

Description of amendment request: This application for amendment to

the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to

change sections 3.7.A.5 and 3.7.F.4 by removing the words ``three

individual underground'' and ``underground'' from the limiting

conditions for operation (LCO) when referring to the emergency diesel

generator (EDG) fuel oil storage tanks (FOSTs).

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 license amendment involve a significant

increase in the probability or consequences of an accident

previously evaluated?

[[Page 29714]]

No. The proposed change would not change the design

configuration or function of the permanently installed EDG FOSTs.

The revision of TS 3.7.A.5 and 3.7.F.4 to remove the descriptive

words ``three individual underground'' and ``underground'' from the

text of the two LCOs is intended as a line item change, to remove

unnecessarily restrictive wording in the TS. While the Standard

Technical Specifications (STS), NUREG-1431, mentions in the Bases

section that ``all outside tanks, pumps, and piping are located

underground'', the specification itself does not contain this

requirement. The intent of this TS change is to allow for, if

acceptable under 10CFR50.59, the potential installation of an

alternate above ground FOST to an EDG if needed to perform repairs/

testing of the permanently installed FOST. This alternate tank would

need to be qualified and have the required capacity to maintain the

associated EDG operable. This potential modification would include

design of the temporary tank to preclude winds loads from a tornadic

event causing the associated EDG to become inoperable. Installation

of this temporary tank would then permit repair work or replacement

of an installed EDG FOST, or subsequent similar work on either of

the other EDG FOSTs, one at a time. The changes to the Bases for

Specification 3.7 are consistent with the change in the LCO

Specification and do not alter the design or functionality of the

existing EDG FOSTs. The revised LCOs are consistent with the STS in

that the FOSTs will no longer be identified as ``three individual

underground''. Control of future modifications to support EDG FOST

work would ensure proper licensing and design basis compliance in

accordance with the change process of 10CFR50.59. The associated

changes of the TS Bases provide clarification regarding the normal

underground configuration of the EDG FOSTs. The proposed TS change

will not reduce the ability of any system, structure, or component

in preventing or mitigating a design basis accident since no plant

features are being altered in conjunction with this change, and

future changes would be evaluated under 10CFR50.59. The description

of the FOSTs, including the fact that they are underground, remains

part of the current licensing basis because it is described in FSAR

[final safety analysis report] section 8.2.

Therefore, the proposed changes to the TS will not result in an

increase in the probability or consequences of any previously

evaluated accidents. The other changes to the TS pages are editorial

only, moving text to different pages.

2. Does the proposed license amendment create the possibility of

a new or different kind of accident from any accident previously

evaluated?

No. The proposed change would not change the design

configuration or function of the permanently installed EDG FOSTs.

The changes to TS 3.7 and its bases in describing the physical

location of the EDG FOSTs will not alter the required design

criteria of these tanks nor their ability to withstand the effects

of a tornado. These changes will not reduce the ability of the EDG's

in meeting their design requirements of providing emergency power

towards mitigating an accident. The intent of these changes is to

permit the potential use of a temporary above ground FOST(s) to

supply the EDGs and to fulfill the intent and requirements of the

present EDG fuel oil storage system while allowing for maintenance

on an EDG FOST. The 10CFR50.59 change process will be used to

determine this potential modification acceptability. The intent of

the temporary configuration of an above ground FOST would be to

maintain the fuel oil system and EDG operable. The associated

changes to the Bases section of TS 3.7 provide additional

clarification of the ``underground'' nature of the EDG FOSTs.

Neither the changes to the LCO in describing the EDG FOSTs (whether

the normal underground tanks or any temporary above ground FOSTs)

nor any changes to the TS Bases (which do not alter the design or

operation of the EDG fuel oil transfer system) will affect the

ability of the EDGs to provide the necessary power for operation of

equipment required for mitigating previously analyzed accident

scenarios. No plant features, or FSAR description of such, are being

altered in conjunction with this change, and future changes would be

evaluated under 10CFR50.59. Therefore, the proposed changes will not

result in an unanalyzed condition and does not create the

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

previously evaluated.

3. Does the proposed amendment involve a significant reduction

in the margin of safety?

No. The proposed changes will not alter any assumptions, initial

conditions, or the results of any accident analyses. The design and

licensing requirements for the EDG fuel oil storage system are

defined in other parts of the IP3 licensing and design basis,

specifically in FSAR section 8.2. Potential modifications supported

by this change would require a subsequent safety evaluation in

accordance with 10 CFR 50.59 regarding the design requirements

(e.g., fire loads, tornadic wind loads, tornado missile criteria,

security, etc.) for an alternate FOST if repairs to present

``underground'' FOSTs are undertaken. The proper design criteria for

the presently installed EDG FOSTs or for potential, alternate EDG

FOSTs will be maintained via present licensing and design basis

requirements and through the 10 CFR 50.59 change process as

required. No plant features are being altered in conjunction with

this change, and future changes would be evaluated under 10 CFR

50.59. Therefore, this proposed license amendment will not involve a

significant reduction in the margin of safety.

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

this review, it appears that the three standards of 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 10601.

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

York, New York 10019.

NRC Section Chief: S. Singh Bajwa.

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

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

Date of amendment request: April 9, 1999.

Description of amendment request: This application for amendment to

the Indian Point 3 Technical Specifications (TSs) proposes to increase

the allowed outage time (AOT) for any one safety injection pump from 24

hours to 72 hours.

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 license amendment involve a significant

increase in the probability or consequences of an accident

previously analyzed?

Response: The proposed 72-hour allowed outage time for any one

safety injection pump does not involve a significant increase in the

probability or consequences of an accident previously analyzed. The

plant Technical Specifications provides allowed outage times for

systems and components to accommodate preventive or corrective

maintenance. A variation in the allowed outage time is not an

accident initiator and thus does not result in a significant

increase in the probability of an accident previously analyzed. The

proposed change provides for an increase in allowed outage time for

any one safety injection pump. The operability of the remaining two

safety injection pumps is required by the Technical Specifications

during this period. The Indian Point 3 High Head Safety Injection

System consists of three safety injection pumps, each capable of

providing 50 percent of the Emergency Core Cooling System [ECCS]

design flow requirement. Therefore, with only one pump inoperable

the remaining two pumps are capable (assuming that no single failure

occurs during the period of the allowed outage time) of mitigating

the consequences of previously analyzed accidents. In addition, a

72-hour allowed outage time for safety injection pumps was evaluated

by the NRC (Reference 3) [NRC Memorandum, R.L. Baer to V. Stello,

``Recommended Interim Revisions to LCOs for ECCS Components,'' dated

December 1, 1975] and generically approved in the Standard Technical

Specifications (Reference 1) [NUREG-1431 ``Standard Technical

Specifications--Westinghouse Plants,'' Revision 1, dated April

1995].

(2) Does the proposed license amendment create the possibility

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

evaluated?

[[Page 29715]]

Response: The proposed 72-hour allowed outage time for any one

safety injection pump does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Changing the allowed outage time is accomplished through

administrative changes, such as changes to plant procedures that

implement Technical Specification requirements for allowed outage

time. This change does not require physical changes to plant systems

or components and also does not involve changes to plant setpoints.

This change also does not affect how the safety injection pumps are

operated under design basis accident conditions. Therefore there are

no changes resulting from the proposed new allowed outage time that

alter system operation or that could create the possibility of a new

or different kind of accident. In addition, a 72-hour allowed outage

time safety injection [pump] was generically approved in the

Standard Technical Specifications.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response: The proposed 72-hour allowed outage time for any one

safety injection pump does not involve a significant reduction in a

margin of safety. With one safety injection pump inoperable, the

remaining two pumps are capable of providing 100% of the fuel

cooling flow assumed for pertinent accident analyses with the

provision that the single-failure assumption is relaxed during the

time period of the allowed outage time. The acceptability of a 72-

hour allowed outage time for ECCS components was established in an

NRC reliability analysis (Reference 3) [NRC Memorandum, R.L. Baer to

V. Stello, ``Recommended Interim Revisions to LCOs for ECCS

Components,'' dated December 1, 1975]. The use of the 72-hour

allowed outage time was generically approved in the Standard

Technical Specifications.

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 10601.

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

York, New York 10019.

NRC Section Chief: S. Singh Bajwa.

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

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

Jersey

Date of amendment request: April 14, 1999.

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 3/4.9.12, ``Fuel Handling Area

Ventilation System (FHAVS),'' to (1) reflect the latest filter testing

standards in the test requirements, (2) add, modify, or delete certain

surveillance test requirements, and (3) clarify the information in the

applicable TS Bases section. The proposed amendments would also make

the TS requirements more consistent with the system design basis.

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.

A Fuel Handling Accident, as described in the Updated Final

Safety Analysis Report (UFSAR) Section 15.4.6, is the design basis

accident considered for establishing system configuration and

performance capability for the FHAVS. This accident is defined as

the dropping of a spent fuel assembly onto the spent fuel rack

resulting in a rupture of the cladding of all the spent fuel rods in

the assembly.

The probability of a fuel handling accident is independent of

the changes proposed in this submittal and it is unaffected by this

submittal. The consequences of a dropped fuel rod are significantly

reduced by pre-aligning the system to its design basis function

prior to moving fuel in the fuel handling building. Pre-aligning the

system eliminates the potential detrimental consequences associated

with a single failure of an active component on the filter train.

The proposed change will not change the way the FHAVS functions to

control the release of radioactive gaseous effluents. Filter testing

is improved by applying more current filter testing requirements to

both Units 1 and 2.

The proposed change will not modify equipment used to store or

move irradiated fuel assemblies, or equipment used to move heavy

loads in the Fuel Handling Building. The proposed new surveillance

will be incorporated into a new or existing procedure.

Therefore, the proposed 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 does not result in any design or physical

configuration changes to the FHAVS, or to the equipment used to

store or move irradiated fuel within the Fuel Handling Building.

Pre-aligning the system to its design basis function prior to moving

fuel in the fuel handling building eliminates the potential

detrimental consequences associated with a single failure of an

active component on the filter train. The system will not be

operated or placed in a configuration that is different from the

configuration that it was designed to operate.

Therefore, the proposed amendment will not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes will ensure that the FHAVS is operated and

tested in accordance to its design basis requirements as specified

in the Salem UFSAR.

The proposed changes will clarify the requirements of the system

to be considered operable to ensure that the FHAVS will perform its

intended safety function in the event of a Fuel Handling Accident.

These changes ensure that the existing margin is maintained and

improved by pre-aligning the system to its accident configuration.

The proposed change does not involve the addition or

modification of plant equipment. It is consistent with the intent of

the existing TS, the design basis of the FHAVS as described in the

UFSAR, and the [Standard Technical Specifications Westinghouse

Plants, NUREG-1431] ITS and associated Bases.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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

West Broadway, Salem, NJ 08079.

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

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

NRC Section Chief: James W. Clifford.

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

Electric Station (CPSES), Units 1 and 2, Somervell County, Texas

Date of amendment request: May 4, 1999.

Brief description of amendments: The proposed license amendments

would change the CPSES Units 1 and 2 Technical Specifications. The

first change revises Surveillance Requirement (SR) 3.8.4.7 to allow the

unrestricted substitution of the modified battery performance discharge

test in lieu of the service discharge test. The second change revises

SRs 3.8.1.7, 3.8.1.12 , 3.8.1.15, and 3.8.1.20 to separate the voltage

and frequency acceptance criteria for the Diesel Generator (DG) start

surveillances into two sets of criteria; those criteria required to be

met within 10 seconds, and those criteria required to be met following

achievement of steady state conditions. The third change corrects

[[Page 29716]]

miscellaneous editorial errors resulting from issuance of Amendment No.

64.

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 consequence of an accident previously evaluated?

(1) Batteries are used to support mitigation of the consequences

of an accident, and are not considered to be an initiator of any

previously analyzed accident. The proposed change would not effect

the design or performance of the batteries. The allowance to perform

the modified performance discharge test in lieu of the service test

at any time is permissible since the test's discharge rate envelopes

the duty cycle of the service test. Therefore, the allowance for

unrestricted substitution of the modified performance discharge test

in lieu of the service discharge test does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

(2) The diesel generators are used to support mitigation of the

consequences of an accident, and are not considered to be an

initiator of any previously analyzed accident. The proposed change

does not affect the accident analysis assumption that the DG reaches

minimum conditions to accept load within 10 seconds. The ability of

the DG to maintain steady state operation within 10 seconds is not

an accident analysis assumption and is primarily used to identify

degradation of governor and voltage regulator performance.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

(3) The editorial changes are non-technical and therefore do not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

different kind of accident from any accident previously evaluated?

(1) The allowance for unrestricted substitution of the modified

performance discharge test in lieu of the service discharge test

does not involve any physical alteration to the plant. No new

failure mechanisms will be introduced and the change does not affect

the ability of the batteries to fulfill their safety-related

function. Therefore, this change does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

(2) The separation of the DG start surveillance criteria into

those criteria required to be met within 10 seconds, and those

criteria required to be met following achievement of steady state

conditions, does not involve any physical alteration to the plant.

No new failure mechanisms will be introduced and the change does not

affect the ability of the DGs to fulfill their safety-related

function. Therefore, this change does not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

(3) The editorial changes are non-technical and therefore 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?

(1) The allowance for unrestricted substitution of the modified

performance discharge test in lieu of the service discharge test

will not alter any accident analysis assumptions, initial

conditions, or results. Consequently, it does not have any effect on

the margin of safety. Therefore, this change does not involve a

significant reduction in a margin of safety.

(2) The proposed change to delete the requirement to demonstrate

that the DG can achieve and maintain steady state operation within

10 seconds is not an accident analysis assumption. The accident

analysis assumption that the DG reaches minimum conditions to accept

load within 10 seconds is preserved. Consequently, it does not have

any effect on the margin of safety. Therefore, this change does not

involve a significant reduction in a margin of safety.

(3) The editorial changes are non-technical and therefore do not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Texas at

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

19497, Arlington, Texas 76019.

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

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

NRC Section Chief: Robert A. Gramm.

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

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

Virginia

Date of amendment request: May 3, 1999.

Description of amendment request: The proposed changes will delete

and/or relocate the additional primary-to-secondary leak rate limits

and enhanced leakage monitoring requirements imposed following the 1987

steam generator tube rupture event.

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:

[O]peration of the North Anna Power Station in accordance with

the proposed Technical Specification changes will not:

Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Eliminating the conservative primary-to-secondary leakage limits

associated with the replaced steam generators and the operability

requirements for the leakage monitoring instrumentation does not

change the operation of the plant. The steam generators will be

operated, inspected, and maintained in the same manner. No new

accident initiators are established as a result of the proposed

changes. Therefore, the probability of occurrence is not increased

for any accident previously evaluated.

Removing the conservative primary-to-secondary leakage limits

associated with the replaced steam generators and the operability

requirements for the leakage monitoring instrumentation does not

change the operation of the plant. Although the conservative leakage

limits are being deleted, the remaining leakage limits will maintain

the dose rate, in the event of a tube rupture, within the analyzed

limits. Therefore, there is no increase in the consequences of any

accident previously analyzed[.]

Create the possibility of a new or different kind of accident

from any accident previously evaluated.

The proposed changes do not affect the operation of the plant.

The steam generators will be operated, inspected, and maintained in

the same manner. There are no modifications to the plant or steam

generators as a result of the change. No new accident or event

initiators are created by the removal of the conservative primary-

to-secondary leakage limits associated with the replaced steam

generators and the operability requirements for the leakage

monitoring instrumentation. Therefore, the proposed changes do not

create the possibility of any accident or malfunction of a different

type.

Involve a significant reduction in the margin of safety as

defined in the bases on any Technical Specifications.

The proposed changes have no effect on any safety analyses

assumptions. The remaining limits maintain primary-to-secondary

leakage within the accident analysis assumptions. The proposed

changes only eliminate overly conservative primary-to-secondary

leakage requirements and the operability and surveillance

requirements for the leakage monitoring system associated with the

replaced steam generators. Therefore, the proposed changes do not

result in a significant reduction in the margin of safety.

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

this review, it appears that the three standards of 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: Mr. Donald P. Irwin, Esq., Hunton and

Williams,

[[Page 29717]]

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

NRC Section Chief: Richard L. Emch, Jr.

Yankee Atomic Electric Co., Docket No. 50-029, Yankee Nuclear Power

Station (YNPS) Franklin County, Massachusetts

Date of amendment request: March 24, 1999.

Description of amendment request: The licensee submitted a request

to delete License Condition 2.C.(10), which states: ``The licensee

shall maintain a Fitness for Duty Program in accordance with the

requirements of 10 CFR Part 26.''

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change is administrative in nature in that it

removes a reference in the YNPS Part 50 License to a regulatory

requirement no longer applicable to a plant which has permanently

ceased power operations and permanently removed fuel from its

reactor vessel. This will permit more cost beneficial use of

available resources with no diminution in the YNPS staff's ability

to maintain the safe operation of the YNPS SFP [spent fuel pool].

The change will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated. Each potential

accident in the YNPS FSAR [final safety analysis report] projects a

maximum release of activity and no prompt mitigation actions. None

of the analyzed scenarios resulted in a situation which could

significantly [a]ffect the public health and safety. Removal of a

regulatory requirement which does not apply to a plant which has

permanently ceased power operations and permanently removed fuel

from its reactor vessel cannot be deemed to involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

any previously evaluated. The proposed change will not modify any

plant systems or components and, therefore will not create the

possibility of a new or different accident from any previously

evaluated.

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

Removal of a regulatory requirement which does not apply to a plant

which has permanently ceased power operations and permanently

removed fuel from its reactor vessel cannot be deemed to involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Greenfield Community College,

1 College Drive, Greenfield, Massachusetts 01301.

Attorney for licensee: Thomas Dignan, Esquire, Ropes and Gray, One

International Place, Boston, Massachusetts 02110-2624.

NRC Section Chief: Michael T. Masnik.

Previously Published Notices of Consideration of Issuance of Amendments

to Facility Operating Licenses, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

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

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

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

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

of the original notice.

Texas Utilities Electric Company, et al., Docket Nos. 50-445 and 50-

446, Comanche Peak Steam Electric Station, Unit Nos. 1 and 2, Somervell

County, Texas

Date of amendment request: February 11, 1999.

Description of amendment request: The proposed amendments would

credit soluble boron in the spent fuel pool water, in the maintenance

of a subcritical condition, and allow an increase in spent fuel storage

from 1291 to 2026 fuel assemblies.

Date of publication of individual notice in Federal Register: May

12, 1999 (64 FR 25522).

Expiration date of individual notice: June 11, 1999.

Local Public Document Room location: University of Texas at

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

19497, Arlington, Texas.

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.

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

Station, Unit No. 1, Shippingport, Pennsylvania

Date of application for amendment: January 17, 1998, as

supplemented by letters dated February 10, 1998, November 9, 1998,

February 8, 1999, and February 26, 1999.

Brief description of amendment: This amendment authorizes changes

to the Beaver Valley Power Station, Unit No. 1 (BVPS-1) Updated Final

Safety Analysis Report (UFSAR). Specifically, the authorized changes to

the UFSAR reflect revisions to the control room radiological dose

calculations for the waste gas system line break accident analysis to

correct a mathematical error discovered in a previous calculation, and

use of more conservative assumptions in the revised analysis.

Date of issuance: May 12, 1999.

[[Page 29718]]

Effective date: As of the date of issuance.

Amendment No: 222.

Facility Operating License No. DPR-66. Amendment approved changes

to the UFSAR.

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

FR 9601).

The February 10, 1998, November 9, 1998, February 8, 1999, and

February 26, 1999, letters provided clarifying information that did not

change the initial proposed no significant hazards consideration

determination or expand the amendment request beyond the scope of the

initial notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 12, 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 No. 50-368, Arkansas Nuclear One, Unit

No. 2, Pope County, Arkansas

Date of application for amendment: August 23, 1996, as supplemented

on April 9, 1999.

Brief description of amendment: The amendment revises Section 5.0,

``Design Features,'' and Section 6.0, ``Administrative Controls,'' of

the Technical Specifications, adopting, for the most part, the format

and content of the NUREG-1432, Revision 1, ``Standard Technical

Specifications [STS] for Combustion Engineering Plants'' for the

changes requested. This amendment also relocates certain portions of

the design features section to other licensee-controlled documents in

accordance with the STS.

Date of issuance: May 19, 1999.

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

within 30 days from the date of issuance: May 19, 1999.

Amendment No: 205.

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

Technical Specifications.

Date of initial notice in Federal Register: October 9, 1996 (61 FR

52965).

The April 9, 1999, letter provided clarifying information that did

not change the scope of the original application and initial proposed

no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 19, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

Entergy Operations, Inc., Docket No. 50-313 and 50-368, Arkansas

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

Date of amendment request: December 19, 1996, as supplemented by

letters dated August 6, 1998, and December 3, 1998.

Brief description of amendments: The amendments change requirements

for the control room ventilation system for both Units 1 and 2.

Date of issuance: May 19, 1999.

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

within 30 days of the date of issuance.

Amendment Nos.: 196 and 206.

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

revised the Technical Specifications.

Date of initial notice in Federal Register: January 29, 1997 (62 FR

4348).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 19, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, Arkansas 72801.

Entergy Operations, Inc., System Energy Resources, Inc., South

Mississippi Electric Power Association, and Entergy Mississippi, Inc.,

Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne

County, Mississippi

Date of application for amendment: January 12, 1999, which

superseded application dated May 31, 1996.

Brief description of amendment: The amendment adds an additional

required action to the Limiting Condition for Operation (LCO) 3.9.1,

``Refueling Equipment Interlocks,'' of the Grand Gulf Technical

Specifications. The additional action will allow an alternative to the

current action for one or more inoperable refueling equipment

interlocks. The current action is to ``suspend in-vessel fuel movement

with equipment associated with the inoperable interlock(s).'' The

alternative action will be to (1) insert a control rod withdrawal

block, and (2) verify all control rods are fully inserted in core cells

containing one or more fuel assemblies. The amendment also revised the

Bases for LCO 3.9.1 actions to describe the alternative action.

Date of issuance: May 7, 1999.

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

within 30 days of issuance.

Amendment No.: 138.

Facility Operating License No. NPF-29: Amendment revises the

Technical Specifications.

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

FR 6695), which superseded original notice of June 16, 1996 (61 FR

31178).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 7, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Judge George W. Armstrong

Library, 220 S. Commerce Street, Natchez, Mississippi 39120.

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

Nuclear Power Station, Unit 1, Ottawa County, Ohio

Date of application for amendment: October 28, 1998, as modified by

letter dated March 19, 1999.

Brief description of amendment: This amendment revises

administrative requirements relating to: TS 6.5.1.6, Station Review

Board Responsibilities; TS 6.8.4.d, Radioactive Effluent Controls

Program; TS 6.10, Records Retention; TS 6.11, Radiation Protection

Program; TS 6.12, High Radiation Area; and TS 6.15, Offsite Dose

Calculation Manual.

Date of issuance: May 19, 1999.

Effective date: May 19, 1999.

Amendment No.: 231.

Facility Operating License No. NPF-3: Amendment revised the

Technical Specifications.

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

FR 64126).

The supplemental information contained clarifying information and

did not change the initial proposed no significant hazards

consideration determination and did not expand the scope of the

application as described in the original Federal Register notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 19, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, OH 43606.

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

Generating Station, Ocean County, New Jersey

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

August 3 (2 letters), September 14, and December 22, 1998.

[[Page 29719]]

Brief description of amendment: The amendment approves the use of a

small amount of containment overpressure to ensure sufficient net

positive suction head for the emergency core cooling system pumps.

Date of Issuance: May 13, 1999.

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

within 30 days.

Amendment No.: 206.

Facility Operating License No. DPR-16. Amendment authorizes changes

to the Updated Final Safety Analysis Report.

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

56250).

The supplemental letters provided additional information that was

within the scope of the original application and did not change the

staff's proposed no significant hazards consideration determination.

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated May 13, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

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

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

Generating Station, Ocean County, New Jersey

Date of application for amendment: November 10, 1998.

Brief description of amendment: The proposed Technical

Specification (TS) change would remove the restriction on the sale or

lease of property within the exclusion area and replace the restriction

with a requirement to retain complete authority to determine and

maintain sufficient control of all activities including the authority

to exclude or remove personnel and property within the minimum

exclusion distance. A TS Bases page for the proposed change is

included. Also included are clarifications and administrative changes

which: (1) clarify TS definition 1.38 to become ``Site Boundary''

rather than the current term ``Exclusion Area'' to be consistent with

the 10 CFR 20.1003 definition for Site Boundary and the 10 CFR 100.3

definition of Exclusion Area, (2) revise the TS definition from

Exclusion Area to Site Boundary in TS 6.8.4(a)(9), and (3) revise and

update the TS Table of Contents for Section I Definitions.

Date of Issuance: May 12, 1999.

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

within 30 days.

Amendment No.: 205.

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

Technical Specifications.

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

FR 66595).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated May 12, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

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

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

Generating Station, Ocean County, New Jersey

Date of application for amendment: February 12, 1999.

Brief description of amendment: The amendment deletes the

organizational chart and related references from the Appendix B

Environmental Technical Specifications (ETS). In addition, the

appearance and format of the ETS have been extensively revised.

Date of Issuance: May 18, 1999.

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

within 30 days from the date of issuance.

Amendment No.: 207.

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

Technical Specifications.

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

17026).

The Commission's related evaluation of this amendment is contained

in a Safety Evaluation dated May 18, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Ocean County Library,

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

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

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

Date of application for amendment: January 18, 1999, as

supplemented February 3 and March 17, 1999.

Brief description of amendment: The amendment removes Technical

Specification (TS) 3/4.6.4.3, ``Containment Systems, Hydrogen Purge

System,'' from the TS and allows downgrading the system to a non-

safety-related system.

Date of issuance: April 12, 1999.

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

60 days from the date of issuance.

Amendment No.: 233.

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

Technical Specifications.

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

FR 6704).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated April 12, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Date of application for amendment: January 18, 1999.

Brief description of amendment: The amendment modifies Technical

Specification 3/4.2.2 to be in accordance with NRC-approved

Westinghouse methodologies for the heat flux hot channel factor--

FQ(Z). In addition, the amendment makes changes to the core

operating limits and the analytical methods used to determine core

operating limits contained in Section 6.9.1.6.a and b, respectively, by

adding, modifying, or deleting references.

Date of issuance: May 10, 1999.

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

30 days from the date of issuance.

Amendment No.: 170.

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

Technical Specifications.

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

FR 6705).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 10, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

ThreeRivers Community-Technical College, 574 New London Turnpike,

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

49 Rope Ferry Road, Waterford, Connecticut

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating

Station, Units 1 and 2, Montgomery County, Pennsylvania.

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

supplemented February 22, 1999.

Brief description of amendments: These amendments revised the

overvoltage, undervoltage, and underfrequency allowable values

[[Page 29720]]

associated with the reactor protection system monitoring channels and

add supporting details to the Technical Specifications Bases 3/4.8.4.

Date of issuance: May 13, 1999.

Effective date: Units 1 and 2, as of date of issuance and shall be

implemented within 30 days.

Amendment Nos.: 134 and 96.

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 18, 1998 (64

FR 64120)

The February 22, 1999, letter provided clarifying information that

did not change the initial proposed no significant hazards

consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 13, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

PECO Energy Company, Docket No. 50-353, Limerick Generating Station,

Unit 2, Montgomery County, Pennsylvania

Date of application for amendment: January 12, 1999, as

supplemented January 29 and March 10, 1999.

Brief description of amendment: This amendment revised Technical

Specifications (TSs) Section 3/4.4.2, ``Safety/Relief Valves,'' and TS

Bases Sections B 3/4.4.2, B 3/4.5.1 and B 3/4.5.2 to increase the

allowable as-found main steam safety relief valve (SRV) code safety

function lift setpoint tolerance from plus or minus 1% to plus or minus

3%. Also, the required number of operable SRVs in operational

conditions 1, 2 and 3 will be increased from 11 to 12.

Date of issuance: May 17, 1999.

Effective date: May 17, 1999.

Amendment No.: 98.

Facility Operating License No. NPF-85. The amendment revises the

Technical Specifications.

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

FR 9194)

The January 29 and March 10, 1999, letters provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination or expand the scope of the original

Federal Register notice.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 17, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

PECO Energy Company, Docket No. 50-353, Limerick Generating Station,

Unit 2, Montgomery County, Pennsylvania

Date of application for amendment: March 11, 1999, as supplemented

April 21, 1999.

Brief description of amendment: The amendment revised the minimum

critical power ratio safety limits and revised the associated Technical

Specification Bases.

Date of issuance: May 14, 1999.

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

prior to restart following completion of the April 1999 refueling

outage.

Amendment No.: 97.

Facility Operating License No. NPF-85. The amendment revised the

Technical Specifications and/or License.

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

17028). The April 21, 1999, letter provided clarifying information that

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 May 14, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

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

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

California

Date of application for amendments: May 22, 1997, as supplemented

by letters dated June 12, 1997, August 28, 1997, January 29, 1998, July

9, 1998, and March 12, 1999.

Brief description of amendments: The amendments authorize revisions

to the licensing basis as described in the Final Safety Analysis Report

(FSAR) Update to incorporate a modification to the Diablo Canyon Power

Plant, Unit Nos. 1 and 2 component cooling water system.

Date of issuance: May 13, 1999.

Effective date: May 13, 1999, and shall be implemented in the next

periodic update to the FSAR Update in accordance with 10 CFR 50.71(e).

Amendment Nos.: Unit 1-134; Unit 2-132.

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

revised the Final Safety Analysis Report Update.

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

40558).

The supplemental letters dated July 9, 1998, and March 12, 1999

provided additional clarifying information, did not expand the scope of

the application as originally noticed, and did not change the staff's

initial no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 13, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: California Polytechnic State

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

Department, San Luis Obispo, California 93407.

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

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

Diego County, California

Date of application for amendments: May 8, 1996, as supplemented

January 13, 1999.

Brief description of amendments: The amendments modified the

technical specifications to allow refueling operation with 20 feet of

water level in the refueling cavity for many operating conditions and

at 12 feet of water level for certain specified conditions. The

amendments also restored a phrase to a note to Limiting Conditions for

Operation for TSs 3.9.4 and 3.9.5 that was inadvertently deleted by

previous amendments.

Date of issuance: May 13, 1999.

Effective date: May 13, 1999, to be implemented within 30 days from

the date of issuance.

Amendment Nos.: Unit 2-153; Unit 3-144.

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

revised the Technical Specifications.

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

14285).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 13, 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 Nos. 50-348 and 50-

364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,

Alabama.

Date of amendments request: April 2, 1999.

Brief Description of amendments: The amendment changes TS 3/4.4.9,

``Specific Activity,'' and the associated bases to increase the limit

associated with dose equivalent iodine-131. The

[[Page 29721]]

steady-state dose equivalent iodine-131 limit would be increased from

0.15Curie/gram to 0.3 Curie/gram and the transient

limit for 80 percent to 100 percent power provided by Technical

Specificaton Figure 3.4-1 will increase 9 Curie/gram to 18

Curie/gram with a corresponding increase in the 0 percent to

80 percent power limits.

Date of issuance: May 10, 1999.

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

implemented within 30 days from the date of issuance.

Amendment Nos.: Unit 1-142; Unit 2-134.

Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise

the Technical Specifications.

Date of initial notice in Federal Register: April 8, 1999 (64 FR

17201).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 10, 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: September 29, 1998.

Brief description of amendments: The amendments authorize the

revision of the South Texas Project updated final safety analysis

report (UFSAR) to incorporate the revised methodology to calculate the

mass and energy release following a postulated large-break loss-of-

coolant accident.

Date of issuance: May 20, 1999.

Effective date: May 20, 1999 Revisions will be incorporated into

the next UFSAR update in accordance with the schedule in 10 CFR

50.71(e).

Amendment Nos.: Unit 1-110; Unit 2-97.

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

authorize the revision of the UFSAR to incorporate the revised

methodology.

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

FR 64123).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 20, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location:

Wharton County Junior College, J. M. Hodges.

Learning Center, 911 Boling Highway, Wharton, Texas 77488.

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

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

Date of amendment request: October 29, 1998, supplemented by letter

dated March 15, 1999.

Brief description of amendments: The amendments relocate the

requirements in Technical Specifications 3/4.7.9 and 6.10.3.l for

snubbers to the Technical Requirements Manual.

Date of issuance: May 17, 1999.

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

30 days from the date of issuance.

Amendment Nos.: Unit 1-109; Unit 2-96.

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

revised the Technical Specifications.

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

FR 69346); renoticed April 7, 1999 (64 FR 17031).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated May 17, 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.

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

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of application for amendment: February 15, 1999.

Brief description of amendment: This amendment revised Technical

Specification Section 6, ``Administrative Controls,'' to reflect

organizational changes, to relocate certain review and audit functions

to the Operational Quality Assurance Program Description, and to

eliminate redundant requirements.

Date of issuance: May 11, 1999.

Effective date: May 11, 1999.

Amendment No.: 145.

Facility Operating License No. DPR-43: Amendment revised the

Technical Specifications.

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

17031).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated May 11, 1999.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Wisconsin,

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

Dated at Rockville, Maryland, this 25th day of May 1999.

For the Nuclear Regulatory Commission.

John A. Zwolinski,

Director, Division of Licensing Project Management, Office of Nuclear

Reactor Regulation.

[FR Doc. 99-13765 Filed 6-1-99; 8:45 am]

BILLING CODE 7590-01-P