Federal Register / Vol. 63, No. 123 / Friday, June 26, 1998 / Notices 34941

qualified to handle radiation

emergencies, and to maintain

arrangements for the transportation of

contaminated individuals to treatment

facilities outside the site boundary.

Paragraph (c) of 10 CFR 70.24 exempts

Part 50 licensees from the requirements

of paragraph (b) of 10 CFR 70.24 for

SNM used or to be used in the reactor.

Paragraph (d) of 10 CFR 70.24 states that

any licensee who believes that there is

good cause why he should be granted an

exemption from all or part of 10 CFR

70.24 may apply to the Commission for

such an exemption and shall specify the

reasons for the relief requested.

III

The SNM that could be assembled

into a critical mass at PBAPS, Units 2

and 3, is in the form of nuclear fuel; the

quantity of SNM other than fuel that is

stored on site in any given location is

small enough to preclude achieving a

critical mass. The Commission’s

technical staff has evaluated the

possibility of an inadvertent criticality

of the nuclear fuel at PBAPS, Units 2

and 3, and has determined that it is

extremely unlikely for such an accident

to occur if the licensee meets the

following seven criteria:

1. Only three new fuel assemblies are

allowed out of a shipping cask or

storage rack at one time.

2. The k-effective does not exceed

0.95, at a 95% probability, 95%

confidence level in the event that the

fresh fuel storage racks are filled with

fuel of the maximum permissible U–235

enrichment and flooded with pure

water.

3. If optimum moderation occurs at

low moderator density, then the keffective does not exceed 0.98, at a 95%

probability, 95% confidence level in the

event that the fresh fuel storage racks

are filled with fuel of the maximum

permissible U–235 enrichment and

flooded with a moderator at the density

corresponding to optimum moderation.

4. The k-effective does not exceed

0.95, at a 95% probability, 95%

confidence level in the event that the

spent fuel storage racks are filled with

fuel of the maximum permissible U–235

enrichment and flooded with pure

water.

5. The quantity of forms of special

nuclear material, other than nuclear

fuel, that are stored on site in any given

area is less than the quantity necessary

for a critical mass.

6. Radiation monitors, as required by

General Design Criterion 63, are

provided in fuel storage and handling

areas to detect excessive radiation levels

and to initiate appropriate safety

actions.

7. The maximum nominal U–235

enrichment is limited to 5.0 weight

percent.

By letter dated March 18, 1998, the

licensee requested an exemption from

10 CFR 70.24.

In this request the licensee addressed

the seven criteria given above. The

Commission’s technical staff has

reviewed the licensee’s submittal and

has determined that PBAPS, Units 2 and

3, meet the applicable criteria. Criteria

2 and 3 are not applicable to PBAPS,

Units 2 and 3, since Technical

Specification Section 4.3.1.2 specifically

states, ‘‘The new fuel storage racks shall

not be used for fuel storage. The new

fuel shall be stored in the spent fuel

storage racks.’’ The reference to General

Design Criterion (GDC) 63 was initially

incorporated to ensure that licensees

receiving an exemption to 10 CFR 70.24

would not erroneously view the

exemption as the basis for removing

from the spent fuel pool area radiation

monitors that were meeting other

monitoring requirements, such as those

contained in GDC 63. However,

Criterion 63 is not applicable to PBAPS

because the units were evaluated against

the draft GDCs current when PBAPS

was licensed rather than the current

GDCs proposed in July 1967. Thus, even

though PBAPS is not required to meet

GDC 63, the staff has determined that it

is extremely unlikely for an inadvertent

criticality to occur in SNM handling and

storage areas at PBAPS, Units 2 and 3.

Additionally, PBAPS, Units 2 and 3,

have area radiation monitors (ARMs)

that meet the requirements of 10 CFR

70.24(a)2, and function as a monitoring

system capable of detecting criticality in

the only area (the refuel floor) where

accidental criticality is possible.

The purpose of the criticality

monitors required by 10 CFR 70.24 is to

ensure that if a criticality were to occur

during the handling of SNM, personnel

would be alerted to that fact and would

take appropriate action. The staff has

determined that it is extremely unlikely

that such an accident could occur. The

low probability of an inadvertent

criticality constitutes good cause for

granting an exemption from the

requirements of 10 CFR 70.24(a).

IV.

The Commission has determined that,

pursuant to 10 CFR 70.14, this

exemption is authorized by law, will not

endanger life or property or the common

defense and security, and is otherwise

in the public interest. Therefore, the

Commission hereby grants PECO Energy

Company, an exemption from the

requirements of 10 CFR 70.24(a) for

Peach Bottom Atomic Power Station,

Units 2 and 3.

Pursuant to 10 CFR 51.32, the

Commission has determined that the

granting of this exemption will have no

significant impact on the environment

(63 FR 33735).

This exemption is effective upon

issuance.

Dated at Rockville, Maryland, this 22nd

day of June 1998.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 98–17095 Filed 6–25–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–346]

Toledo Edison Company, et al.; (DavisBesse Nuclear Power Station, Unit 1);

Confirmatory Order Modifying License,

Effective Immediately

I

Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company (the

Licensees) are the holders of Facility

Operating License No. NPF–3, which

authorizes operation of the Davis-Besse

Nuclear Power Station, Unit 1, located

in Ottawa County, Ohio.

II

The staff of the U.S. Nuclear

Regulatory Commission (NRC) has been

concerned that Thermo-Lag 330–1 fire

barrier systems installed by licensees

may not provide the level of fire

endurance intended and that licensees

using Thermo-Lag 330–1 fire barriers

may not be meeting regulatory

requirements. During the 1992 to 1994

timeframe, the NRC staff issued Generic

Letter (GL) 92–08, ‘‘Thermo-Lag 330–1

Fire Barriers,’’ and subsequent requests

for additional information that asked

licensees to submit plans and schedules

for resolving the Thermo-Lag issue. The

NRC staff has obtained and reviewed all

such corrective plans and schedules.

The staff is concerned that some

licensees may not be making adequate

progress toward resolving the plantspecific issues, and that some

implementation schedules may be either

too tenuous or too protracted. For

example, several licensees informed the

NRC staff that their completion dates

had slipped by 6 months to as much as

3 years. The NRC staff has met with

licensees of plants that have completion

action scheduled beyond 1997 to
34942 Federal Register / Vol. 63, No. 123 / Friday, June 26, 1998 / Notices

discuss the progress of the licensees’

corrective actions and the extent of

licensee management attention

regarding completion of Thermo-Lag

corrective actions. In addition, the NRC

staff discussed with licensees the

possibility of accelerating their

completion schedules.

The NRC staff met with the Licensees

for Davis-Besse on April 3, 1997. At this

meeting, the NRC staff reviewed the

schedule of Thermo-Lag corrective

actions described in the Licensees’

submittals to the NRC dated February

20, April 24, June 26, and November 5,

1996, as documented in the NRC

meeting summary dated April 16, 1997.

On the basis of the information

submitted by the Licensees (including

an additional letter dated September 10,

1997), the NRC staff has concluded that

the schedules presented are reasonable.

This conclusion is based on (1) the

amount of installed Thermo-Lag; (2) the

complexity of the plant-specific fire

barrier configurations and issues; and

(3) the need to perform certain plant

modifications during outages as

opposed to those that can be performed

while the plant is at power. In order to

remove compensatory measures such as

fire watches, it has been determined that

resolution of the Thermo-Lag corrective

actions by the Licensees must be

completed in accordance with their

current schedule. By letter dated May 4,

1998, the NRC staff notified the

Licensees of its plan to incorporate their

schedule commitment into a

requirement by issuance of an order and

requested consent from the Licensees.

By letter dated June 11, 1998, the

Licensees provided their consent to

issuance of a Confirmatory Order.

III

The Licensees’ commitment as set

forth in their letter of June 11, 1998, is

acceptable and is necessary for the NRC

to conclude that public health and

safety are reasonably assured. To

preclude any schedule delay and to

ensure public health and safety, the

NRC staff has determined that the

Licensees’ commitment in their June 11,

1998, letter be confirmed by this Order.

The Licensees have agreed to this

action. On this basis, and the Licensees’

consent, this Order is immediately

effective upon issuance.

IV.

Accordingly, pursuant to sections

103, 161b, 161i, 161o, 182, and 186 of

the Atomic Energy Act of 1954, as

amended, and the Commission’s

regulations in 10 CFR 2.202 and 10 CFR

Part 50, it is hereby ordered, effective

immediately, that

The Toledo Edison Company, Centerior

Service Company, and The Cleveland

Electric Illuminating Company (the licensees)

shall complete final implementation of

Thermo-Lag 330–1 fire barrier corrective

actions at the Davis-Besse Nuclear Power

Station, Unit No. 1, by December 31, 1998,

as described in the licensees’ submittals to

the NRC dated February 20, 1996, April 24,

1996, June 26, 1996, November 5, 1996, and

September 10, 1997, and as presented at the

licensees’ meeting with the NRC staff on

April 3, 1997, as documented in the NRC

meeting summary dated April 16, 1997.

The Director, Office of Nuclear

Reactor Regulation, may relax or

rescind, in writing, any provisions of

this Confirmatory Order upon a showing

by the Licensees of good cause.

V

Any person adversely affected by this

Confirmatory Order, other than the

Licensees, may request a hearing within

20 days of its issuance. Where good

cause is shown, consideration will be

given to extending the time to request a

hearing. A request for extension of time

must be made in writing to the Director,

Office of Nuclear Reactor Regulation,

U.S. Nuclear Regulatory Commission,

Washington, D.C. 20555–0001, and must

include a statement of good cause for

the extension. Any request for a hearing

shall be submitted to the Secretary, U.S.

Nuclear Regulatory Commission,

Attention: Rulemakings and

Adjudications Staff, Washington, D.C.

20555–0001. Copies of the hearing

request shall also be sent to the Director,

Office of Nuclear Reactor Regulation,

U.S. Nuclear Regulatory Commission,

Washington, D.C. 20555–0001, to the

Deputy Assistant General Counsel for

Enforcement at the same address, to the

Regional Administrator, NRC Region III,

801 Warrenville Road, Lisle, Illinois

60532–4351, and to the Licensees. If

such a person requests a hearing, that

person shall set forth with particularity

the manner in which his/her interest is

adversely affected by this Order and

shall address criteria set forth in 10 CFR

2.714(d).

If a hearing is requested by a person

whose interest is adversely affected, the

Commission will issue an Order

designating the time and place of any

such hearing. If a hearing is held, the

issue to be considered at such hearing

shall be whether this Confirmatory

Order should be sustained.

In the absence of any request for

hearing, or written approval of an

extension of time in which to request a

hearing, the provisions specified in

Section IV above shall be final 20 days

from the date of this Order without

further order or proceedings. If an

extension of time for requesting a

hearing has been approved, the

provisions specified in Section IV shall

be final when the extension expires if a

hearing request has not been received.

An answer or a request for hearing shall

not stay the immediate effectiveness of

this Order.

Dated at Rockville, Maryland this 22nd day

of June 1998.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 98–17098 Filed 6–25–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 40–8502]

Cogema Mining, Inc.; Environmental

Statements; Availability, etc

AGENCY: Nuclear Regulatory

Commission.

ACTION: Final Finding of No Significant

Impact and Notice of Opportunity for

Hearing.

SUMMARY: The U.S. Nuclear Regulatory

Commission (NRC) proposes to renew

NRC Source Material License SUA–1341

to authorize the licensee, COGEMA

Mining, Inc. (COGEMA), to continue the

commercial operation of its in-situ leach

(ISL) uranium mines and processing

facilities, located in Campbell and

Johnson Counties, Wyoming. This

license currently authorizes COGEMA

to receive, acquire, possess, and transfer

uranium at its Irigaray and Christensen

Ranch Facilities, which are located

approximately 10 miles northeast of

Sussex, Wyoming, and 30 miles northnortheast of Midwest, Wyoming,

respectively. An Environmental

Assessment (EA) was performed by the

NRC staff in support of its review of

COGEMA’s license renewal request, in

accordance with the requirements of 10

CFR Part 51. The conclusion of the

Environmental Assessment is a Finding

of No Significant Impact (FONSI) for the

proposed licensing action.

FOR FURTHER INFORMATION CONTACT: Ms.

Janet Lambert, Uranium Recovery

Branch, Mail Stop TWFN 7-J9, Division

of Waste Management, Office of Nuclear

Material Safety and Safeguards, U.S.

Nuclear Regulatory Commission,

Washington, DC 20555. Telephone 301/

415–6710. E-mail: JAL@NRC.GOV

SUPPLEMENTARY INFORMATION:

Background

The Irigaray Project was licensed for

commercial operation in August 1978,