7800 Federal Register / Vol. 60, No. 27 / Thursday, February 9, 1995 / Notices

related to 4.5.1.b.3 in a previous TS

change.

Environmental Impacts to the Proposed

Action

The proposed changes will provide

additional time for operator action in a

boron dilution event to assure that there

is at least 15 minutes between the time

to boron dilution alarm assuming an

alarm penalty of 1.3 and the time to

criticality for Modes 1 through 5 and 30

minutes for Mode 6 for operator action.

The Commission has completed its

evaluation of the proposed TS changes

and concludes that the combination of

the shutdown margin increases and the

lower credited boron dilution alarm

setpoint assuming an alarm penalty

factor of 1.3 will provide assurance that

the criteria for operator action will be

met. In addition, the neutron sources

will be moved further away from the

excore detectors for the Cycle 19 startup

(approximately March 1995). This will

provide additional margin in the alarm

setpoint as the need for any penalty

factor will be significantly reduced or

completely eliminated. In addition, the

staff agrees that the change in references

in TS 3.1.2.2, 3.1.2.4, and 3.1.2.6, and

Surveillance Specifications 4.1.2.3.1

and 4.1.2.4.1 are editorial in nature.

The proposed TS change will not

increase the probability or consequences

of accidents, no changes are being made

in the types of any effluents that may be

released offiste, and there is no

significant increase in the allowable

individual or cumulative occupational

radiation exposure. Accordingly, the

Commission concludes that there are no

significant radiological environmental

impacts associated with this proposed

TS amendment.

With regard to potential

nonradiological impacts, the proposed

amendment does involve features

located entirely within the restricted

area as defined in 10 CFR Part 20. It

does not affect nonradiological plant

effluents and has no other

environmental impact. Accordingly, the

Commission concludes that there are no

significant nonradiological

environmental impacts associated with

the proposed amendment.

Alternatives to the Proposed Action

Since the Commission has concluded

there is no measurable environmental

impact associated with the proposed

amendment, any alternatives with equal

or greater environmental impact need

not be evaluated. As an alternative to

the proposed action, the staff considered

denial of the proposed action. Denial of

application would result in no change

in current environmental impacts. The

environmental impacts of the proposed

action and the alternative action are

similar.

Alternative Use of Resources

This action does not involve the use

of resources not considered previously

in the Final Environmental Statement

for the Haddam Neck Plant.

Agencies and Persons Consulted

In accordance with its stated policy,

the staff consulted with the Connecticut

State official regarding the

environmental impact of the proposed

action. The State official has no

comments.

Finding of No Significant Impact

Based upon the environmental

assessment, the Commission concludes

that the proposed action will not have

a significant effect on the quality of the

human environment. Accordingly, the

Commission has determined not to

prepare an environmental impact

statement for the proposed amendment.

For further details with respect to this

proposed action, see the licensee’s letter

dated September 7, 1994, 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 located at the

Russell Library, 123 Broad Street,

Middletown, CT 06547.

Dated at Rockville, Maryland, this 3rd day

of February 1995.

For the Nuclear Regulatory Commission.

Phillip F. McKee,

Director, Project Directorate I–4, Division of

Reactor Projects—I/II, Office of Nuclear

Reactor Regulation.

[FR Doc. 95–3232 Filed 2–8–95; 8:45 am]

BILLING CODE 7590–01–M

Northeast Nuclear Energy Company;

Millstone Nuclear Power Station, Unit

No. 3; Environmental Assessment and

Finding of No Significant Impact

The U.S. Nuclear Regulatory

Commission (the Commission) is

considering issuance of an amendment

to Facility Operating License No. NPF–

49, issued to Northeast Nuclear Energy

Company (the licensee), for operation of

the Millstone Nuclear Power Station,

Unit No. 3, located in New London

County, Connecticut.

Environmental Assessment

Identification of the Proposed Action

The proposed action would revise

Technical Specification (TS) 3.5.2.a by

granting a one-time extension of the

allowable Residual Heat Removal (RHR)

pump outage time for mechanical seal

replacement and related modifications

from 72 hours to 120 hours. This

exception would only be used one time

per pump and expire on April 30, 1995.

The amendment would clearly define

the times in which each RHR pump and

associated RHR heat exchanger must be

restored to an operable state.

The proposed action is in accordance

with the licensee’s application for

amendment dated August 16, 1994, as

supplemented by letter dated January

10, 1995.

The Need for the Proposed Action

The proposed action would reduce

the potential for an unnecessary plant

shutdown, thus, eliminating a source of

unnecessary challenges to the plant’s

safety systems.

Environmental Impacts of the Proposed

Action

The Commission has completed its

evaluation of the proposed action and

concludes that one-time extension of the

RHR pump outage time from 72 hours

to 120 hours to acceptable.

The change will not increase the

probability or consequences of

accidents, no changes are being made in

the types of any effluents that may be

released offsite, and there is no

significant increase in the allowable

individual or cumulative occupational

radiation exposure. Accordingly, the

Commission concludes that there are no

significant radiological environmental

impacts associated with the proposed

action.

With regard to potential

nonradiological impacts, the proposed

action does involve features located

entirely within the restricted area as

defined in 10 CFR part 20. It does not

affect nonradiological plant effluents

and has no other environmental impact.

Accordingly, the Commission concludes

that there are no significant

nonradiological environmental impacts

associated with the proposed action.

Alternatives to the Proposed Action

Since the Commission has concluded

there is no measurable environmental

impact associated with the proposed

action, any alternatives with equal or

greater environmental impact need not

be evaluated. As an alternative to the

proposed action, the staff considered

denial of the proposed action. Denial of

the application would result in no

change in current environmental

impacts. The environmental impacts of

the proposed action and the alternative

action are similar.
Federal Register / Vol. 60, No. 27 / Thursday, February 9, 1995 / Notices 7801

Alternative Use of Resources

This action does not involve the use

of any resources not previously

considered in the Final Environmental

Statement for the Millstone Nuclear

Power Station, Unit No. 3.

Agencies and Persons Consulted

In accordance with its stated policy,

the staff consulted with the Connecticut

State official regarding the

environmental impact of the proposed

action. The State official had no

comments.

Finding of No Significant Impact

Based upon the environmental

assessment, the Commission concludes

that the proposed action will not have

a significant effect on the quality of the

human environment. Accordingly, the

Commission has determined not to

prepare an environmental impact

statement for the proposed action.

For further details with respect to the

proposed action, see the licensee’s letter

dated August 16, 1994, as supplemented

by letter dated January 10, 1995, which

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 room located at the

Learning Resource Center, Three Rivers

Community-Technical College, Thames

Valley Campus, 574 New London

Turnpike, Norwich, CT 06360.

Dated at Rockville, Maryland, this 3rd day

of February 1995.

For the Nuclear Regulatory Commission

Phillip F. McKee,

Director, Project Directorate I–4, Division of

Reactor Projects—I/II, Office of Nuclear

Reactor Regulation.

[FR Doc. 95–3233 Filed 2–8–95; 8:45 am]

BILLING CODE 7590–01–M

[Docket No. 50–413]

Duke Power Company, et al.; Notice of

Consideration of Issuance of

Amendment to Facility Operating

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing

The U.S. Nuclear Regulatory

Commission (the Commission) is

considering issuance of an amendment

to Facility Operating License No. NPF–

35 issued to Duke Power Company (the

licensee) for operation of the Catawba

Nuclear Station, Unit 1, located in York

County, South Carolina.

The proposed amendment request

would propose the renewal for Catawba

Unit 1 Cycle 9 operation of the steam

generator tube inspection bobbin probe

voltage-based interim plugging criteria

that had been previously approved for

Cycle 8. Approval of this amendment

will preclude unnecessary plugging or

repairing tubes by sleeving due to the

occurrence of outer diameter initiated

stress corrosion cracking (ODSCC) at the

tube support plate elevations in the

Catawba Unit 1 steam generators. The

interim plugging criteria approved for

Cycle 8 and contained in the draft

Generic Letter 94–XX, ‘‘Voltage-Based

Repair Criteria for the Repair of

Westinghouse Steam Generator Tubes

Affected by Outside Diameter Stress

Corrosion Cracking,’’ can be

summarized as follows:

Flaw indications with a bobbin coil voltage

less than or equal to 1.0 volt can remain in

service without further action. For flaw

indications in excess of 1.0 volt but less than

2.7 volts, the tube can remain in service

provided an RPC inspection of the indication

does not detect ODSCC or any other

degradation mode. Crack indications above

2.7 volts will be plugged or repaired by

sleeving, and do not require RPC

confirmation.

This amendment request reflects the

‘‘Requested Actions: for a licensee that

chooses to implement a steam generator

tube interim plugging criteria, as stated

in the draft NRC Generic Letter, 94–XX

‘‘Voltage-Based Repair Criteria for the

Repair of Westinghouse Steam

Generator Tubes Affected by Outside

Diameter Stress Corrosion Cracking.’’

The changes being proposed to the

Technical Specification (TS) do not alter

the interim plugging criteria currently

stated in the TS which was approved

and utilized during Cycle 8. The

primary change to the TS is to

incorporate the guidance of draft

Generic Letter 94–XX, ‘‘Voltage-Based

Repair Criteria for the Repair of

Westinghouse Steam Generator Tubes

Affected by Outside Diameter Stress

Corrosion Cracking,’’ which will allow

removal of the cycle-specific limitation

currently in the TS.

Before issuance of the proposed

license amendment, the Commission

will have made findings required by the

Atomic Energy Act of 1954, as amended

(the Act) and the Commission’s

regulations.

The Commission has made a

proposed determination that the

amendment request involves 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. 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 Catawba Unit 1 in

accordance with the proposed license

amendment does not involve a significant

increase in the probability or consequences

of an accident previously evaluated.

A single tube rupture is not anticipated

during operation of Catawba Unit 1. Based on

the existing data base, the limiting RG

[Regulatory Guide] 1.121 criterion for tube

burst capability of 3 times normal operating

differential is satisfied with 3⁄4′′ diameter

tubing with bobbin coil indications with

signal amplitudes less than 4.54 volts,

regardless of the indicated depth

measurement. This structural limit is based

on a lower 95% prediction bound of the data

and using LTL material properties. A 1.0 volt

plugging criteria compares favorably with the

structural limit considering the previously

calculated growth rates for ODSCC within the

Catawba Unit 1 steam generators. Assuming

a voltage increase of 0.4 volts, and adding a

14% NDE uncertainty of 0.14 volts (90%

cumulative probability) to the interim

plugging criteria [IPC] of 1.0 volt results in

an EOC [end-of-cycle] voltage of

approximately 1.6 volts. This end of cycle

voltage compares favorably with the

Structural Limit of 4.54 volts. The

applicability of assumed growth rates for

each cycle of operation will be confirmed

prior to return to power of Catawba Unit 1.

A similar structural margin is anticipated for

subsequent cycles.

In addition, for an EOC voltage structural

limit of 4.54 volts, applying the 40% growth

allowance and the 14% NDE uncertainty

results in a margin between the structural

limit and the alternate repair limit (2.7 volts),

which is well within the structural limit.

This repair limit will be applied for IPC

implementation to repair bobbin indications

greater than 2.7 volts independent of RPC

confirmation of the indication.

Concerning SLB [steamline break] leakage

in support of implementation to the interim

plugging criteria, it will be determined

whether the distribution of cracking

indications at the tube support plate

intersections at the end of a cycle are

projected to be such that primary to

secondary leakage would result in site

boundary doses within the pertinent 10 CFR

100 limits. The SLB leakage rate calculation

methodology * * * will be used to calculate

End of Cycle SLB leakage. Based on EOC 8

projections, it is calculated that leakage

during a postulated SLB event at the EOC 8

will be limited to approximately 1.61 gpm

which is shown to result in acceptable dose

consequences. [An] SLB leakage of 17.5 gpm

in the faulted loop results in dose

consequences which are less than the

pertinent 10 CFR 100 limits. Similar results

are expected for subsequent cycles and

confirmation of leak rates will be performed

prior to placing the [s]team generators in

service.