33782 Federal Register / Vol. 61, No. 126 / Friday, June 28, 1996 / Notices

is precluded through compliance with

the Farley Technical Specifications,

including reactivity requirements (e.g.,

shutdown margins, limits on control rod

movement), instrumentation

requirements (e.g., reactor power and

radiation monitors), and controls on

refueling operations (e.g., control rod

interlocks and source range monitor

requirements). In addition, the

operators’ continuous attention directed

toward instruments monitoring behavior

of the nuclear fuel in the reactor assures

that the facility is operated in such a

manner as to preclude inadvertent

criticality. Finally, since access to the

fuel in the reactor vessel is not

physically possible while in use and is

procedurally controlled during

refueling, there are no concerns

associated with loss or diversion of the

fuel.

SNM as nuclear fuel is stored in one

of two locations—the spent fuel pool or

the new fuel storage area. The spent fuel

pool is used to store irradiated fuel

under water after its discharge from the

reactor. The pool is designed to store the

fuel in a geometric array that precludes

criticality. In addition, existing

Technical Specification limits on keff are

maintained less than or equal to 0.95,

even in the event of a fuel handling

accident.

The new fuel storage area is used to

receive and store new fuel in a dry

condition upon arrival on site and prior

to loading in the reactor. The new fuel

storage area is designed to store new

fuel in a geometric array that precludes

criticality. In addition, existing safety

evaluations demonstrate that keff is

maintained less than or equal to 0.95

when the new fuel racks are fully

loaded and dry or flooded with

unborated water and less than or equal

to 0.98 for optimum moderation

conditions (e.g., because of the presence

of aqueous foam or mist) or in the event

of a fuel handling accident.

Fresh fuel is shipped in a plastic

wrap. In some cases the fuel is stored in

the new fuel storage racks with the

plastic wrap in place and in other cases

the plastic wrap is removed prior to

storage. In all cases where fuel is stored

with the plastic wrap in place, the wrap

either cannot hold water due to its

design or it is rendered incapable of

holding water prior to fuel storage.

Therefore, there is no concern that the

plastic wrap used as part of fresh fuel

storage will hold water from flooding

from overhead sources. Additionally, as

discussed above, the new fuel storage

racks have been analyzed for a

postulated flooded condition and the

results showed that keff is maintained

less than or equal to 0.95.

Both irradiated and unirradiated fuel

is moved to and from the reactor vessel,

and the spent fuel pool to accommodate

refueling operations. Also, unirradiated

fuel can be moved to and from the new

fuel storage area. In addition,

movements of fuel into the facility and

within the reactor vessel or within the

spent fuel pool occur. In all cases, fuel

movements are procedurally controlled

and designed to preclude conditions

involving criticality concerns.

Moreover, previous accident analyses

have demonstrated that a fuel handling

accident (i.e., a dropped fuel element)

will not create conditions which exceed

design specification. In addition, the

Technical Specifications specifically

address the refueling operations and

limit the handling of fuel to ensure

against an accidental criticality and to

preclude certain movements over the

spent fuel pool and the rector vessel.

In summary, exemptions from the

requirements of 10 CFR Part 70, Section

70.24 approved by the NRC in

connection with the SNM licenses for

Farley Units 1 and 2 were based upon

NRC’s finding that the inherent features

associated with the storage and

inspection of unirradiated fuel

established good cause for granting the

exemption and that granting such an

exemption at this time will not

endanger public life or property or the

common defense and security and is

otherwise in the public interest. The

training provided to all personnel

involved in fuel handling operations,

the administrative controls, the

Technical Specifications on new and

spent fuel handling and storage, and the

design of the new and spent fuel storage

racks in place preclude inadvertent or

accidental criticality. Since the

facilities, storage, and inspection and

procedures currently in place are

consistent with those in place at the

time the exemptions were granted in

connection with the SNM licenses, an

exemption from 10 CFR 70.24 is

appropriate.

The proposed exemption will not

affect radiological plant effluents nor

cause any significant occupational

exposures. Only a small amount, if any,

of radioactive waste is generated during

the receipt and handling of new fuel

(e.g., smear papers or contaminated

packaging material). The amount of

waste would not be changed by the

exemption.

With regard to potential

nonradiological impacts, the proposed

exemption involves systems located

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

that there is no measurable

environmental impact associated with

the proposed action, any alternatives

with equal or greater environmental

impact need not be evaluated. The

principal alternative would be to deny

the requested exemption. 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 any resources not previously

considered in the Final Environmental

Statement related to the operation of

Joseph M. Farley Nuclear Plant, Units 1

and 2, dated June 1972.

Agencies and Persons Consulted

In accordance with its stated policy,

on June 14, 1996, the staff consulted

with the Alabama State official, Mr. Kirk

Whatley, of the Alabama Department of

Public Health, 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 May 31, 1996, 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

Houston-Love Memorial Library, 212 W.

Burdeshaw Street, Post Office Box 1369,

Dolthan, Alabama.

Dated at Rockville, Maryland, this 20th day

of June 1996.

For The Nuclear Regulatory Commission.

Byron L. Siegel,

Senior Project Manager, Project Directorate

II–2, Division of Reactor Projects—I/II, Office

of Nuclear Reactor Regulation.

[FR Doc. 96–16554 Filed 6–27–96; 8:45 am]

BILLING CODE 7590–01–P
Federal Register / Vol. 61, No. 126 / Friday, June 28, 1996 / Notices 33783

Advisory Committee on Reactor

Safeguards; Meeting of the ACRS

Subcommittee on Westinghouse

Standard Plant Designs; Notice of

Meeting

The ACRS Subcommittee on

Westinghouse Standard Plant Designs

will hold a meeting on July 19, 1996,

Room T–2B3, 11545 Rockville Pike,

Rockville, Maryland.

The entire meeting will be open to

public attendance with the exception of

a portion that may be closed to discuss

Westinghouse proprietary information

pursuant to 5 U.S.C. 552b(c)(4).

The agenda for the subject meeting

shall be as follows:

Friday, July 19, 1996—8:30 a.m. until

the conclusion of business

The Subcommittee will discuss

SECY–96–128, ‘‘Policy and Key

Technical Issues Pertaining to the

Westinghouse AP600 Standardized

Passive Reactor Design,’’ dated June 12,

1996, which contains proposed staff

positions on three policy issues:

Prevention and Mitigation of Severe

Accidents, Post-72-Hour Actions, and

External Reactor Vessel Cooling, as well

as the status of resolution of seven key

technical issues. The purpose of this

meeting is to gather information,

analyze relevant issues and facts, and to

formulate proposed positions and

actions, as appropriate, for deliberation

by the full Committee.

Oral statements may be presented by

members of the public with the

concurrence of the Subcommittee

Chairman; written statements will be

accepted and made available to the

Committee. Electronic recordings will

be permitted only during those portions

of the meeting that are open to the

public, and questions may be asked only

by members of the Subcommittee, its

consultants, and staff. Persons desiring

to make oral statements should notify

the cognizant ACRS staff engineer

named below five days prior to the

meeting, if possible, so that appropriate

arrangements can be made.

During the initial portion of the

meeting, the Subcommittee, along with

any of its consultants who may be

present, may exchange preliminary

views regarding matters to be

considered during the balance of the

meeting.

The Subcommittee will then hear

presentations by and hold discussions

with representatives of the NRC staff,

Westinghouse Electric Corporation, and

other interested persons regarding this

review.

Further information regarding topics

to be discussed, whether the meeting

has been cancelled or rescheduled, the

Chairman’s ruling on requests for the

opportunity to present oral statements

and the time allotted therefor can be

obtained by contacting the cognizant

ACRS staff engineer, Mr. Noel F. Dudley

(telephone 301/415–6888) between 7:30

a.m. and 4:15 p.m. (EDT). Persons

planning to attend this meeting are

urged to contact the above named

individual one or two working days

prior to the meeting to be advised of any

potential changes to the agenda, etc.,

that may have occurred.

Dated: June 24, 1996.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 96–16552 Filed 6–27–96; 8:45 am]

BILLING CODE 7590–01–P

Advisory Committee on Reactor

Safeguards, Joint Meeting of the ACRS

Subcommittees on Probabilistic Risk

Assessment and on Plant Operations

The ACRS Subcommittees on

Probabilistic Risk Assessment and on

Plant Operations will hold a joint

meeting on July 17–18, 1996, Room T–

2B3, 11545 Rockville Pike, Rockville,

Maryland.

The entire meeting will be open to

public attendance.

The agenda for the subject meeting

shall be as follows:

Wednesday, July 17, 1996—8:30 a.m.

until the conclusion of business

The Subcommittees will discuss riskbased analysis of reactor operating

experience.

Thursday, July 18, 1996—8:30 a.m. until

the conclusion of business

The Subcommittees will discuss the

issues identified in the Staff

Requirements Memoranda dated May 15

and June 11, 1996, including the role of

performance-based regulation in the

PRA Implementation Plan, plantspecific application of safety goals,

requirement for risk neutrality versus

the allowance for an acceptable increase

in risk, risk-informed inservice testing

(IST) and inservice inspection (ISI)

requirements, and methods for judging

the acceptability and unacceptability of

assumptions and models used in

performing PRAs. The Subcommittees

will also discuss the pilot applications

for risk-informed and performancebased regulations. The purpose of this

meeting is to gather information,

analyze relevant issues and facts, and to

formulate proposed positions and

actions, as appropriate, for deliberation

by the full Committee.

Oral statements may be presented by

members of the public with the

concurrence of the Subcommittee

Chairman; written statements will be

accepted and made available to the

Committee. Electronic recordings will

be permitted only during those portions

of the meeting that are open to the

public, and questions may be asked only

by members of the Subcommittees, their

consultants, and staff. Persons desiring

to make oral statements should notify

the cognizant ACRS staff engineer

named below five days prior to the

meeting, if possible, so that appropriate

arrangements can be made.

During the initial portion of the

meeting, the Subcommittees, along with

any of their consultants who may be

present, may exchange preliminary

views regarding matters to be

considered during the balance of the

meeting.

The Subcommittees will then hear

presentations by and hold discussions

with representatives of the NRC staff, its

consultants, representatives of the

Nuclear Energy Institute, and other

interested persons regarding this review.

Further information regarding topics

to be discussed, whether the meeting

has been cancelled or rescheduled, the

Chairman’s ruling on requests for the

opportunity to present oral statements

and the time allotted therefor can be

obtained by contacting the cognizant

ACRS staff engineer, Mr. Michael T.

Markley (telephone 301/415–6885)

between 7:30 a.m. and 4:15 p.m. (EDT).

Persons planning to attend this meeting

are urged to contact the above named

individual one or two working days

prior to the meeting to be advised of any

potential changes to the agenda, etc.,

that may have occurred.

Dated: June 20, 1996.

Sam Duraiswamy,

Chief, Nuclear Reactors Branch.

[FR Doc. 96–16553 Filed 6–27–96; 8:45 am]

BILLING CODE 7590–01–P

UNITED STATES POSTAL SERVICE

BOARD OF GOVERNORS

Sunshine Act Meetings; Notice of Vote

to Amend Agenda

At the June 3, 1996, meeting of the

Board of Governors, noticed in the

Federal Register on May 14, 1996 (61

FR 24341), and May 23, 1996, (61 FR

25928), the members voted

unanimously to add to its agenda

consideration of an officer change and

officers’ compensation, and that no

earlier public announcement of the new

item on the agenda was possible.

The Board determined that discussion

of the matters would likely disclose

information relating to internal

personnel practices.