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

Federal RegisterAug 11, 1999

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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 July 17, 1999, through July 30, 1999. The

last biweekly notice was published on July 28, 1999 (64 FR 40903).

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-

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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 September 10, 1999, the licensee may file a request for a

hearing with respect to issuance of the amendment to the subject

facility operating license and any person whose interest may be

affected by this proceeding and who wishes to participate as a party in

the proceeding must file a written request for a hearing and a petition

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

to intervene shall be filed in accordance with the Commission's ``Rules

of Practice for Domestic Licensing Proceedings'' in 10 CFR part 2.

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

Carolina

Date of amendment request: July 9, 1999.

Description of amendment request: The proposed amendment would

revise Harris Nuclear Plant (HNP) Technical Specification (TS) 3/4.2.2,

``Heat Flux Hot Channel Factor--FQ(Z),'' TS 3/4.2.3, ``RCS

Flow Rate And Nuclear Enthalpy Rise Hot Channel Factor,'' TS 3/4.2.5,

``DNB Parameters,'' an associated note in TS Table 2.2-1, and

associated Bases. Specifically, the proposed amendment would: (1)

Remove the allowance for reduced power operation for reduced Reactor

Coolant System (RCS) flow rate conditions; (2) separate the

requirements for F delta H and RCS flow rate in the format prescribed

by NUREG-1431, Revision 1, ``Standard Technical Specifications,

Westinghouse Plants,'' dated April 1995; and, (3) implement the

guidance of NUREG-1431, Revision 1, and NRC Generic Letter (GL) 88-16,

dated October 4, 1988 for TS 3/4.2.2, TS 3/4.2.3, TS 3/4.2.5 and

associated Bases by removing cycle specific parameters and placing that

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information into the Core Operating Limits Report (COLR).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed amendment will not introduce any new equipment or

require existing equipment to function different from that

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

TS.

As described in HNP TS Bases, the limits on heat flux hot

channel factor, RCS flow rate, and enthalpy rise hot channel factor

ensure that: (1) the design limits on peak local power density and

minimum DNBR [departure from nucleate boiling ratio] are not

exceeded and (2) in the event of a LOCA the peak fuel clad

temperature will not exceed the 2200 degree Fahrenheit ECCS

[emergency core cooling system] acceptance limit.

Removing the allowance for reduced power operation for reduced

RCS flow conditions is more restrictive than that currently allowed

by TS. Power Distribution Limiting Conditions for Operation for heat

flux hot channel factor and enthalpy rise hot channel factor are not

affected by this change. Therefore, the consequences of an accident

will not increase because of this change. Power Distribution limits

place administrative restrictions on reactor core parameters and as

such do not initiate nor mitigate accidents.

Power Distribution limits at HNP are developed using NRC

approved methodologies. Changing power distribution limits to be

consistent with NUREG-1431, Revision 1 will not increase the

probability or consequences of an accident that has been previously

evaluated.

Relocating cycle specific information from TS to the COLR will

not impact the ability of structures, systems, or components to

mitigate accidents. Future changes to relocated requirements in the

COLR will be submitted to the NRC for review in accordance with HNP

TS Section 6.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The proposed amendment will not introduce any new equipment or

require existing equipment to function different from that

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

TS. The changes are consistent with NUREG-1431, Revision 1 and the

Commission's Final Policy Statement on Technical Specification

improvements. The proposed amendment will not create any new

accident scenarios, because the change does not introduce any new

single failures, adverse equipment or material interactions, or

release paths.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The LCO limit for RCS flow rate at 100.0% reactor power has not

changed. The previous capability to operate with reduced RCS flow

rate has been eliminated. This aspect of the proposed change is more

restrictive than current plant TS in that continued reactor

operation greater than 5% is not allowed if RCS flow rate is less

than the LCO limit at 100% power.

Changes to TS 3/4.2.2, TS 3/4.2.3, TS

3/4.2.5 and associated Bases are in accordance with NUREG-1431,

Revision 1. The completion times for TS Actions are acceptable

because the plant is not allowed to remain in an unacceptable

condition for an extended period of time. Sufficient time to reduce

reactor power in an orderly manner or perform other required actions

is also provided. The surveillance intervals established by NUREG-

1431, Revision 1 have been determined to be adequate for monitoring

the change in power distribution.

Relocating cycle specific information from HNP TS to the COLR is

in accordance with NRC GL 88-16. HNP does not intend to alter the

methodologies for any parameter limit calculation as a result of

this change. The proposed change is in accordance with the plant

safety analysis. Therefore, the proposed change does not involve a

significant reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

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

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

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

Raleigh, North Carolina 27602.

NRC Section Chief: Sheri R. Peterson.

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

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

Carolina

Date of amendment request: July 9, 1999.

Description of amendment request: The proposed amendment would

relocate Harris Nuclear Plant (HNP) Technical Specification (TS) 3/

4.3.3.3, ``Seismic Instrumentation,'' TS

3/4.3.3.4, ``Meteorological Instrumentation,'' TS 3/4.3.3.9, ``Metal

Impact Monitoring System,'' and TS

3/4.3.3.11, ``Explosive Gas Monitoring Instrumentation,'' to plant

procedure PLP-114, ``Relocated Technical Specifications and Design

Basis Requirements.'' The proposed change is in accordance with

guidance provided by NRC Generic Letter 95-10, ``Relocation of Selected

Technical Specification Requirements Related to Instrumentation.''

Changes to relocated requirements would be performed in accordance with

10 CFR 50.59. Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Seismic Instrumentation, Meteorological Instrumentation, Metal

Impact Monitoring System, and Explosive Gas Monitoring

Instrumentation are not accident initiating components as described

in the Final Safety Analysis Report. Seismic Instrumentation,

Meteorological Instrumentation, Metal Impact Monitoring System, and

Explosive Gas Monitoring Instrumentation are not accident mitigating

components. There are no modifications being made to plant systems

as a result of this change. Additionally, there are no changes being

made to the way in which systems are being operated as a result of

this change. Therefore, the proposed change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

Seismic Instrumentation, Meteorological Instrumentation, Metal

Impact Monitoring System, and Explosive Gas Monitoring

Instrumentation are not accident initiating components as described

in the Final Safety Analysis Report (FSAR). The proposed change

relocates the TS requirements for Seismic Instrumentation,

Meteorological Instrumentation, Metal Impact Monitoring System, and

Explosive Gas Monitoring Instrumentation to plant procedure PLP-114.

Plant systems and components are not modified as a result of this

change. Future changes in these systems will be controlled in

accordance with 10 CFR 50.59.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed change to Seismic Instrumentation, Meteorological

Instrumentation, Metal Impact Monitoring System, and Explosive Gas

Monitoring Instrumentation does not affect any of the

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parameters that relate to the margin of safety as described in the

Bases of the TS or the FSAR. Accordingly, NRC Acceptance Limits are

not affected by this change. The proposed change relocates the TS

requirements for Seismic Instrumentation, Meteorological

Instrumentation, Metal Impact Monitoring System, and Explosive Gas

Monitoring Instrumentation to plant procedure PLP-114. Plant systems

and components are not modified as a result of this change. Future

changes in these systems will be controlled in accordance with 10

CFR 50.59. Generic Letter 95-10 states that the staff has concluded

that these provisions are not related to dominant contributors to

plant risk.

Therefore, the proposed change does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

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

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

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

Raleigh, North Carolina 27602.

NRC Section Chief: Sheri R. Peterson.

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

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

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos.

1 and 2, Will County, Illinois.

Date of amendment request: June 30, 1999

Description of amendment request: The proposed amendment would

clarify that the source of DC electrical power required for a unit in

Mode 5 or 6 or during the movement of irradiated fuel assemblies may be

cross-tied to the opposite unit. An administrative change would also

delete reference to AT&T batteries since all AT&T batteries have been

replaced with Charter Power Systems, Inc. (C&D) batteries. The

amendment would also remove the Allowed Outage Time (AOT) extension

approved for Braidwood Station by Amendment No. 99. The activity

addressed by Amendment No. 99 is complete and the extension no longer

applies.

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 change involve a significant increase in

the probability or consequences of an accident previously evaluated?

The proposed change will allow one DC bus on a shutdown unit to

be supplied via the DC bus cross-tie to the opposite unit. The other

DC bus on the shutdown unit will at all times be required to be

fully operable, supplied by the associated battery and charger, and

the associated cross-ties open. The DC electrical system is not

considered an initiator of any accident previously evaluated, and

therefore the probability of a previously analyzed accident is

unchanged.

The consequences of a previously analyzed event are dependent on

the initial conditions assumed for the analysis, the availability

and successful functioning of the equipment assumed to operated in

response to the analyzed event, and the setpoints at which these

actions are initiated. Sufficient equipment remains available to

mitigate the consequences of previously analyzed events. The Updated

Final Safety Analysis Report (UFSAR) section 8.3.2.1.1 clearly

allows operation with the DC cross-tie closed on one DC bus between

a unit that is operating and a unit that is shutdown, or between two

shutdown units, in the manner proposed by this amendment. The TS in

effect prior to the implementation of the Improved TS also allowed

operation in the manner proposed by this amendment. If DC buses are

cross-tied due to an inoperable DC source on a shutdown unit, both

the previous TS and the change proposed by this amendment limit the

time in this condition to seven days, and if the inoperable source

is a battery, the current on the cross-tie is limited to 200 amps.

These actions protect both the operating unit, and the shutdown

unit. If a shutdown unit's DC bus is cross-tied to an operating

unit's DC bus due to an inoperable charger on the operating unit,

both the previous TS and the change proposed by this amendment limit

the time in this condition to 24 hours. The limitations imposed by

both the previous TS and the change proposed by this amendment

ensure that operation in this configuration is within the design

bases of the plant. Thus the consequences of accidents previously

analyzed are unchanged between the previous TS and the change

proposed by this amendment. In the worst case scenario, assuming a

single failure, one DC bus on the shutdown unit will always be

operable, and the ability to mitigate the consequences of any

accident previously analyzed is preserved.

The change to delete all references in the Braidwood TS to AT&T

batteries and the AOT extension granted under TS Amendment Number 99

is administrative only, and has no impact on the probability or

consequences of accidents previously evaluated.

Therefore this proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical change to the

plant. No new equipment is being introduced, and installed equipment

is not being operated in a new or different manner. There is no

change being made to the parameters within which the plant is

operated. There are no setpoints affected by this change at which

protective or mitigative actions are initiated. This change will not

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

the function demands on credited equipment be changed. No alteration

in the procedures which ensure the plant remains within analyzed

limits in being proposed, and no change is being made to the

procedures relied upon to respond to an off-normal event. As such,

no new failure modes are being introduced. The change does not alter

assumptions made in the safety analysis and licensing basis.

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

different kind of accident from any accident previously evaluated.

The change to delete all references in the Braidwood TS to AT&T

batteries and the AOT extension granted under TS Amendment Number 99

is administrative only, and cannot create the possibility of a new

or different kind of accident.

Therefore, the proposed changes do not create the possibility of

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

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

margin of safety?

The margin of safety is established through equipment design,

operating parameters, and the setpoints at which automatic actions

are initiated. Sufficient equipment remains available to actuate

upon demand for the purpose of mitigating an analyzed event. The

proposed change, which will allow one DC bus on a shutdown unit to

be supplied via the DC bus cross-tie to the opposite unit, is

acceptable because of the limitations imposed on operation in this

configuration, and because the other DC bus on the shutdown unit

will at all times be required to be fully operable, supplied by the

associated battery and charger, and the associated cross-ties open.

The TS in effect prior to the implementation of the Improved TS

allowed operation in the manner proposed by this amendment. In the

worst case scenario, assuming a single failure, one DC bus on the

shutdown unit will always be operable. Thus, there is no detrimental

impact on any equipment design parameter, and the plant will still

be required to operate within prescribed limits. Therefore, the

change does not reduce the margin of safety.

The change to delete all references in the Braidwood TS to AT&T

batteries and the AOT extension granted under TS Amendment Number 99

is administrative only, and 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 10 CFR 50.92(c) are

satisfied. Therefore, the NRC staff

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proposes to determine that the requested amendments involve no

significant hazards consideration.

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

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

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

Wilmington, Illinois 60481.

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

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

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

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

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

Date of amendment request: May 3, 1999.

Description of amendment request: The proposed amendments would

relocate Technical Specifications (TS) Section 3/4.6.I to the Updated

Final Safety Analysis Report (UFSAR). TS Section 3/4.6.I contains

reactor coolant chemistry limiting conditions for operation (LCO) and

surveillance requirements (SR) for conductivity, chloride

concentration, and pH.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes simplify the TS, meet regulatory

requirements for relocated TS's, and implement the recommendations

of the NRC Final Policy Statement on TS improvements. The Chemistry

requirements will be relocated to the Updated Final Safety Analysis

Report (UFSAR) and to applicable station procedures. Future changes

to these requirements will be controlled by 10 CFR 50.59. The

proposed changes are administrative in nature and do not involve any

modification to any plant equipment or affect plant operation.

Therefore, the proposed changes do not involve a significant

increase in the probability or consequences of any previously

evaluated accident.

Consequently, this proposed amendment does not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The proposed changes are administrative in nature, do not

involve any physical alterations to any plant equipment, and cause

no change in the method by which any safety related system performs

its function. Therefore, this proposed TS amendment will not create

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

accident previously evaluated.

Does the change involve a significant reduction in a margin of

safety?

The proposed amendment represents the relocation of current

requirements, which are based on generic guidance or previously

approved provisions for other stations. The proposed changes are

administrative in nature and do not adversely affect existing plant

safety margins or the reliability of the equipment assumed to

operate in the safety analysis. The proposed changes have been

evaluated and found to be acceptable for use at Dresden Nuclear

Power Station. Since the proposed changes are administrative in

nature, and are based on NRC accepted provisions which have been

adopted at other nuclear facilities, and maintain the necessary

levels of system reliability, the proposed changes do not involve a

significant reduction in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

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

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

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket No. 50-373, LaSalle County Station,

Unit 1, LaSalle County, Illinois

Date of amendment request: July 7, 1999.

Description of amendment request: The proposed amendments would (1)

revise Technical Specification Section 2.1, Safety Limits, to reflect a

change to the LaSalle, Unit 1, Minimum Critical Power Ratio Safety

Limit; and (2) revise Technical Specification Section 6.6.A.6 to add an

NRC-approved Siemens Power Corporation methodology to the list of

topical reports used to determine the core operating limits.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The probability of an evaluated accident is derived from the

probabilities of the individual precursors to that accident. The

consequences of an evaluated accident are determined by the

operability of plant systems designed to mitigate those

consequences. Limits have been established consistent with NRC-

approved methods to ensure that fuel performance during normal,

transient, and accident conditions is acceptable. These changes do

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

any fuel performance limits.

Changing the MCPR Safety Limit for LaSalle Unit 1 will not

increase the probability or the consequences of an accident

previously evaluated. This change implements the MCPR Safety Limit

resulting from the SPC ANFB critical power correlation methodology

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

cycle, cycle specific MCPR Safety Limit calculations will be

performed, consistent with SPC's approved methodology, to confirm

the appropriateness of the MCPR Safety Limit. Additionally,

operational MCPR limits will be applied that will ensure the MCPR

Safety Limit is not violated during all modes of operation and

anticipated operational occurrences. The MCPR Safety Limit ensures

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

experience boiling transition. Therefore the probability or

consequences of an accident will not increase.

Adding EMF-85-74, Revision 0, Supplement 1 (P)(A) and Supplement

2 (P)(A) to Section 6 does not increase the probability or

consequences of an accident previously evaluated. The NRC-approved

burnup extension for RODEX2A applications has been demonstrated to

meet all applicable design criteria. Therefore adding this

methodology to Technical Specification Section 6 does not increase

the probability or consequences of an accident previously evaluated

Therefore, this proposed amendment does not involve a

significant increase in the probability or consequences of any

accident previously evaluated.

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

kind of accident from any accident previously evaluated?

Creation of the possibility of a new or different kind of

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

that accident. New accident precursors may be created by

modifications to the plant configuration, including changes in

allowable modes of operation. This Technical Specification submittal

does not involve any modifications to the plant configuration or

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

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

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

new or different kind of accident from any accident previously

evaluated.

Changing the MCPR Safety Limit does not create the possibility

of a new accident from any accident previously evaluated. This

[[Page 43769]]

change does not alter or add any new equipment or change modes of

operation. The MCPR Safety Limit is established to ensure that 99.9%

of the rods avoid boiling transition.

The MCPR Safety Limit is changing for LaSalle Unit 1 to support

Cycle 9 operation. This change does not introduce any physical

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

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

are created that are different from any accident previously

evaluated.

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

(P)(A) and Supplement 2 (P)(A)) does not create the possibility of a

new accident from an accident previously evaluated. This change does

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

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

alter the processes used to operate the plant, or change allowable

modes of operation. Therefore, no new accidents are created that are

different from any accident previously evaluated.

Therefore, the proposed changes do not create the possibility of

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

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

of safety?

Changing the MCPR Safety Limit for LaSalle Unit 1 will not

involve any reduction in margin of safety. The MCPR Safety Limit

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

rods are calculated to be in boiling transition. The proposed

Technical Specification amendment request reflects the MCPR Safety

Limit results from evaluations by SPC using NRC-approved

methodology.

The revised MCPR Safety Limit will ensure the same level of fuel

protection. Additionally, operational limits will be established

based on the proposed MCPR Safety Limit to ensure that the MCPR

Safety Limit is not violated during all modes of operation including

anticipated operation[al] occurrences. This will ensure that the

fuel design safety criterion of more than 99.9% of the fuel rods

avoiding transition boiling during normal operation as well as

during an anticipated operational occurrence is met.

The addition of EMF-85-74, Revision 0, Supplement 1 (P)(A) and

Supplement 2 (P)(A) to Section 6 does not decrease the margin of

safety. The burnup limit extension for RODEX2A applications has been

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

extension demonstrates that all applicable design criteria are met.

Therefore, since the burnup extension is acceptable and within the

design criteria, using the approved burnup extension will not affect

the margin of safety.

Therefore, these changes do not involve a significant reduction

in the margin of safety.

Therefore, based upon the above evaluation, ComEd has concluded

that these changes involve no significant hazards consideration.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library, 815

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

Illinois 61348-9692.

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

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

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

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

Date of amendment request: November 9, 1998, as supplemented on

July 7, 1999.

Description of amendment request: The proposed amendments would

revise Technical Specification Table 3.3.3-2, ``Emergency Core Cooling

System Actuation Instrumentation Setpoints,'' to modify the degraded

voltage second level undervoltage relay setpoint and allowable value.

These proposed amendments were originally noticed on January 13, 1999

(64 FR 2245), and are being renoticed to include the revised setpoints

that were included in the July 7, 1999, supplement.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The setpoint change does not change the logic or function of the

degraded voltage protection circuits as described in the UFSAR

[Updated Final Safety Analysis Report] Section 8.2.3. They also do

not reduce the reliability of these circuits. The increase in the

degraded voltage protection circuit setpoint is conservative

compared to the existing setpoint. There is no change as a result of

this amendment to the underlying accident and transient analyses

that support operations of LaSalle County Station. Inadvertent or

spurious operation of the degraded voltage protection function will

initiate loading of the safe shutdown loads on the diesel generators

and is not assumed to initiate an accident. The proposed degraded

voltage setpoints are low enough to prevent spurious actuations

given the expected offsite grid voltages. After implementation of

this amendment, no operator actions are required for equipment

operations in response to degraded voltage conditions.

This change does not affect the initiators or precursors of any

accident previously evaluated. This change will not increase the

likelihood that a transient initiating event will occur because

transients are initiated by equipment malfunction and/or

catastrophic system failure.

The consequences of accidents previously evaluated are not

increased. The proposed change does not affect the required level of

availability of systems required to mitigate the accidents

considered in the analyses. The proposed changes will ensure that

the Class 1E equipment will be capable of starting and operating

during a design basis accident with degraded offsite grid voltage.

The increase in the level of confidence is the result of more

rigorous methodology used to determine limiting Class 1E bus

voltages at the minimum expected offsite AC voltage. These

calculations demonstrate that the degraded voltage relays will not

actuate following a block start of the electrical loads that are

automatically actuated by or as a consequence of the LOCA [loss-of-

coolant accident] signal if the switchyard voltage remains above 352

kV.

If the grid voltage drops below 352 kV, then the analytical

limit of 3814 volts for proper operation of class 1E loads connected

to each 4.16 kV Class 1E bus is assured by transfer to the

respective onsite power sources (Emergency Diesel Generators (EDGs))

by the degraded voltage logic.

Therefore this proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

accident from any accident previously evaluated because:

Setpoint methodology established the bases to ensure that, with

known errors, the relays will detect degraded voltage conditions and

transfer safety loads to the EDGs at a voltage level adequate to

ensure proper safety equipment performance and to prevent equipment

damage.

The trip setpoint of greater than or equal to 3863 volts and

less than or equal to 3877 volts and the allowable value of greater

than or equal to 3814 volts and less than or equal to 3900 volts,

include adequate tolerance to calibrate the relay trip units while

ensuring that the Class 1E bus voltage will remain above the

analytical limits.

These setpoint changes will ensure that adequate voltages will

be available for the continuous operation of safety-related

equipment required to function during a LOCA. These proposed changes

will also ensure that adequate voltages will be available for

starting any Class 1E equipment.

The proposed degraded voltage setpoint change does not change

the design of the degraded voltage protection system or its function

to protect against degraded offsite power. Actuation of the degraded

voltage protection system will initiate a sequence of events that

will start the EDG for the associated Class 1E bus, strip loads from

the Class 1E bus, open all feed breakers to the Class 1E bus, close

the Emergency feed breaker (thus energizing the Class 1E bus

[[Page 43770]]

from the respective EDG), and initiate starting of the Safe Shutdown

equipment supplied by the Class 1E bus.

Since the scope of this change does not affect the operation of

auxiliary power system or any actions necessary to mitigate the

consequences of accidents or achieve safe shutdown, the change does

not involve a new or different accident scenario.

Therefore, these proposed changes do not create the possibility

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

evaluated.

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

because:

The proposed amendment will allow the degraded voltage setpoint

to be conservatively established based on new engineering

calculations which consider the lowest expected offsite grid voltage

and operation of required Class 1E equipment under design basis

accident loading conditions.

The proposed degraded voltage setpoints will ensure that

adequate Class 1E bus voltage will be available to support starting

and operation of required Class 1E loads. The proposed setpoint

includes instrument error to ensure that the lowest possible voltage

will not be lower than the degraded voltage analytical limits.

Additionally, the proposed setpoints are low enough to prevent

spurious actuations due to expected fluctuations in the grid

voltage. The new setpoints are also set with margin to the minimum

Class 1E bus voltage, which is based on a minimum grid voltage of

352 kV, which is less than the expected grid voltage of 354 kV. The

proposed changes will provide an increase in the level of protection

that currently exists and will ensure the margin of safety is

adequately maintained.

Therefore, these changes do not involve a significant reduction

in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library, 815

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

Illinois 61348-9692.

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

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

767, Chicago, Illinois 60690-0767.

NRC Section Chief: Anthony J. Mendiola.

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

Plant, Unit 1, Monroe County, Michigan

Date of amendment request: April 20, 1999 (Reference NRC-99-0035).

Description of amendment request: The proposed amendment will

revise the Technical Specifications by deleting Specification D.3.c.

Specification D.3.c requires the licensee to perform weekly

observations of the nitrogen cover gas pressure within the sodium

storage tanks located in the Sodium Building Complex. Removing this

surveillance requirement would allow the licensee to remove the

nitrogen cover gas system from service for these sodium storage tanks.

This action is necessary for the licensee to begin work on removing the

remaining residual sodium from these tanks. The licensee also requested

an editorial change to delete the words ``STORAGE TANK'' from the title

of Specification D.

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 using the standards in 10 CFR 50.92(c). The licensee's

analysis is presented below:

(1) The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

Removing the primary cover gas supply from the storage tanks

will not significantly increase the probability of an accident

occurring as long as the probability of an uncontrolled water

reaction with residual sodium is not significantly increased. This

is ensured by sealing the storage tanks after the nitrogen cover gas

system is removed except when controlled activities such as sampling

are performed. The consequences of an accident would not be affected

by removing the nitrogen cover gas supply from service as the

previously analyzed primary sodium accident already involves release

of all the radioactive material in the primary sodium. Removing the

cover gas will not increase the amount of radioactive material

available to be released.

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

or different accident from any previously evaluated.

A sodium accident has been previously evaluated. No other type

of accident could be caused by removing the primary sodium tanks

cover gas or opening the tanks since no other system or mode of

operation of any other system will be affected.

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

in a margin of safety.

Currently, only a small amount of residual sodium remains in the

primary sodium storage tanks. Some of this residual sodium may have

been converted to sodium carbonate. This conversion of sodium to

sodium carbonate would have left even less sodium remaining in these

tanks. The cover gas is a good precaution, especially for tanks

sitting unattended for many years. It prevents moisture from

intruding into the tanks and reacting with the sodium residues. It

also prevents oxygen from entering these tanks and reacting with any

hydrogen formed from reactions of water and sodium. Discontinuing

the use of cover gas slightly reduces the margin of safety, but not

significantly. Removing the cover gas does not, in itself, introduce

water into the tank in an uncontrolled manner. Even if slight

amounts of moisture from humidity in the air enter these tanks over

the next year or two, until the sodium is removed while the tanks

are either opened or sealed, the volume of each tank (15,000

gallons) is large enough that the tank should be able to dissipate

any small reactions that could occur. The design pressure for the

primary sodium storage tanks is from vacuum to 50 pounds per square

inch based on the vendor's drawing.

Even if sufficient water entered the tank, generated hydrogen,

and sufficient oxygen entered the tank to cause a reaction that

released the contents of the tank, there would be no significant

release of radioactivity from the tank. The release of all residual

primary sodium would result in concentration levels well below the

values in 10 CFR 20, Appendix B, Table II for releases to

unrestricted areas. Since there is less sodium in the primary sodium

storage tanks than in the secondary sodium storage tanks, potential

hazard consequences of releasing the contents of a primary sodium

tank are bounded by the hypothetical secondary sodium scenario

evaluated in the Fermi 1 Safety Analysis Report. For these reasons,

the proposed change does not involve a significant reduction in the

margin of safety.

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, NRC staff proposes to determine that the

amendment request involves no significant hazards consideration.

Local Public Document Room location: Monroe County Library System,

3700 South Custer Road, Monroe, Michigan 48161.

Attorney for licensee: John Flynn, Esquire, Detroit Edison Company,

2000 Second Avenue, Detroit, Michigan 48226.

NRC Branch Chief: Larry W. Camper.

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

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

Date of amendment request: July 22, 1998, supplemented by October

22, 1998, January 28, May 6 and June 24, 1999.

Description of amendment request: By the referenced submittals the

licensee requested the Catawba Technical Specifications be changed to

permit the licensee's planned use of fuel supplied by Westinghouse,

which has different design characteristics from the fuel currently in

use. The staff has previously published two Notices of Consideration of

Issuance of Amendments and Proposed No Significant Hazards

Consideration of Issuance of Amendments. The first notice, dated

November 18, 1998 (63 FR 64108), covers the submittals dated July

[[Page 43771]]

22 and October 22, 1998. The second notice, dated May 19, 1999 (64 FR

27317), covers the submittal dated May 6, 1999. The June 24, 1999,

submittal actually requested an amendment separate from that described

above, but nevertheless conveyed a revised proposed Figure 2.1.1-1,

``Reactor Core Safety Limits--Four Loops in Operation'', superseding

what was originally proposed in the licensee's previous submittals.

Hence, this Notice only covers the revised proposed Figure 2.1.1-1. The

Notices referenced above are unaffected.

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 for the June 24, 1999, submittal. The staff has reviewed

the licensee's analysis and has performed its own analysis as follows:

First Standard

No. The proposed changes to Figure 2.1.1-1 will not affect the

safety function and will not involve any change to the design or

operation of any plant system or component. The revised Figure 2.1.1-1

restricts reactor coolant flow to within previously analyzed

temperature and pressure conditions. Therefore, no accident

probabilities or consequences will be impacted.

Second Standard

No. The proposed changes will not lead to any hardware or operating

procedure change. Hence, no new equipment failure modes or accidents

from those previously evaluated will be created.

Third Standard

No. Margin of safety is associated with confidence in the design

and operation of the plant; specifically, the ability of the fission

product barriers to perform their design functions during and following

an accident. The proposed changes to Figure 2.1.1-1 do not involve any

change to plant design, operation, or analysis. Thus, the margin of

safety previously analyzed and evaluated is maintained.

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

CFR 50.92(c) are satisfied for the proposed change to Figure 2.1.1-1.

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

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

Black Street, Rock Hill, South Carolina.

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

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

North Carolina.

NRC Section Chief: Richard L. Emch, Jr.

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

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

Date of amendment request: June 24, 1999.

Description of amendment request: The proposed amendments would

change the Technical Specifications (TS) as follows: (1) Revise Figure

2.1.1-1, ``Reactor Core Safety Limits--Four Loops in Operation,'' which

defines the current limits of reactor coolant system (RCS) flow under

different combinations of pressure and temperature; (2) revise the

Actions associated with Limiting Condition of Operation (LCO) 3.4.1 and

Table 3.4.1-1 to reflect the updated assumptions for reactor coolant

flow, temperature and pressure; and (3) delete Figure 3.4.1-1, ``RCS

Total Flow Rate Versus Rated Thermal Power--Four Loops in Operation,''

since these requirements are being relocated to LOC 3.4.1 and Table

3.4.1-1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration for the June 24, 1999, submittal, which is presented

below:

First Standard

No component modification, system realignment, or change in

operating procedure will occur which could affect the probability of

any accident or transient. The increase in RCS total flow rate limit

will not change the probability of actuation of any Engineered

Safety Feature or other device. In order to provide more margin in

the core design limits and allow more flexibility for future cycle-

specific core design, the analyses that establish these limits were

reanalyzed at the proposed TS minimum RCS total flow rate limit. The

impact of the power/flow tradeoff is determined for each reanalyzed

event either by qualitative evaluation or by explicit reanalysis.

An increase in the Technical Specification minimum RCS total

flow rate limit and the revised power/flow tradeoff will not

adversely affect the steady-state or transient analyses documented

in Chapters 3, 4, 6, and 15 of the McGuire and Catawba Nuclear

Station UFSARs [Updated Final Safety Analysis Reports]. The reduced

RCS low flow reactor trip setpoint and allowable value will not

increase the consequences of the partial loss of forced reactor

coolant flow and reactor coolant pump shaft seizure accidents. In

these transient reanalyses, the minimum DNBR and peak primary system

pressure acceptance criteria are not adversely affected. Therefore,

the proposed changes will not involve an increase in the probability

or consequences of an accident previously evaluated.

Second Standard

No component modification, system realignment, or change in

operating procedure will occur which could create the possibility of

a new or different kind of accident. As described in Attachment 3,

the proposed increase in Technical Specification minimum RCS total

flow rate limit and revised power/flow tradeoff will not adversely

affect the steady-state or transient analyses documented in Chapters

3, 4, 6, and 15 of the McGuire and Catawba Nuclear Station UFSARs.

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

new or different kind of accident from any accident previously

evaluated.

Third Standard

These amendments will not involve a significant reduction in a

margin of safety. As described in Attachment 3, the increase in

minimum RCS total flow rate limit and revised power/flow tradeoff

will not adversely affect the steady-state or transient analyses

documented in Chapters 3, 4, 6, and 15 of the McGuire and Catawba

Nuclear Station UFSARs. DNBR, fuel clad intergrity, reactor vessel

integrity and containment integrity will not be adversely affected

by the proposed changes. Therefore, the proposed changes will not

involve any 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: York County Library, 138 East

Black Street, Rock Hill, South Carolina.

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

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

North Carolina.

NRC Section Chief: Richard L. Emch, Jr.

Duke Energy Corporation, et al., Docket Nos. 50-369 and 50-370, McGuire

Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: July 22, 1998, supplemented by October

22, 1998, January 28, May 6 and June 24, 1999.

Description of amendment request: By the referenced submittals the

licensee requested the McGuire Technical Specifications be changed to

permit the licensee's planned use of fuel supplied by Westinghouse,

which has different design characteristics from the fuel currently in

use. The staff has previously published two Notices of

[[Page 43772]]

Consideration of Issuance of Amendments and Proposed No Significant

Hazards Consideration of Issuance of Amendments. The first notice,

dated December 16, 1998 (63 FR 69338), covers the submittals dated July

22 and October 22, 1998. The second notice, dated May 19, 1999 (64 FR

35202), covers the submittal dated May 6, 1999. The June 24, 1999,

submittal actually requested an amendment separate from that described

above, but nevertheless conveyed a revised proposed Figure 2.1.1-1,

``Reactor Core Safety Limits--Four Loops in Operation,'' superseding

what was originally proposed in the licensee's previous submittals.

Hence this Notice only covers the revised proposed Figure 2.1.1-1. The

Notices referenced above are unaffected.

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 for the June 24, 1999, submittal. The staff has reviewed

the licensee's analysis, and has performed its own analysis as follows:

First Standard

No. The proposed changes to Figure 2.1.1-1 will not affect the

safety function and will not involve any change to the design or

operation of any plant system or component. The revised Figure 2.1.1-1

restricts reactor coolant flow to within previously analyzed

temperature and pressure conditions. Therefore, no accident

probabilities or consequences will be impacted.

Second Standard

No. The proposed changes would not lead to any hardware or

operating procedure change. Hence, no new equipment failure modes or

accidents from those previously evaluated will be created.

Third Standard

No. Margin of safety is associated with confidence in the design

and operation of the plant; specifically, the ability of the fission

product barriers to perform their design functions during and following

an accident. The proposed changes to Figure 2.1.1-1 do not involve any

change to plant design, operation or analysis. Thus, the margin of

safety previously analyzed and evaluated is maintained.

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

CFR 50.92(c) are satisfied for the proposed change to Figure 2.1.1-1.

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina.

Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,

422 South Church Street, Charlotte, North Carolina.

NRC Section Chief: Richard L. Emch, Jr.

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear

Station, Units 1 and 2, Mecklenburg County, North Carolina

Date of amendment request: June 24, 1999.

Description of amendment request: The proposed amendments would

change the Technical Specifications (TS) as follows: (1) Revise Figure

2.1.1-1, ``Reactor Core Safety Limits--Four Loops in Operation,'' which

defines the current limits of reactor coolant system (RCS) flow under

different combinations of pressure and temperature; (2) revise Table

3.3.1-1 to provide values for the trip setpoint and allowable value for

RCS Flow-Low; (3) revise Table 3.3.1-1 to make a typographical

correction for T, the nominal T-average at Rated Thermal Power; (4)

revise the Actions associated with Limiting Condition of Operation

(LCO) 3.4.1 and Table 3.4.1-1 to reflect the updated assumptions for

reactor coolant flow, temperature and pressure; and (5) delete Figure

3.4.1-1, ``RCS Total Flow Rate Versus Rated Thermal Power--Four Loops

in Operation,'' since these requirements are being relocated to LCO

3.4.1 and Table 3.4.1-1.