34490 Federal Register / Vol. 63, No. 121 /Wednesday, June 24, 1998 / Notices

Road, Wilmington, North Carolina

28403–3297.

Dated at Rockville, Maryland, this 17th day

of June 1998.

For the Nuclear Regulatory Commission.

David C. Trimble,

Project Manager, Project Directorate II–1,

Division of Reactor Projects—I/II , Office of

Nuclear Reactor Regulation.

[FR Doc. 98–16743 Filed 6–23–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–327 and 50–328]

Tennessee Valley Authority (Sequoyah

Nuclear Plant, Units 1 and 2);

Confirmatory Order Modifying License

Effective Immediately

I

Tennessee Valley Authority (TVA, or

the Licensee) is the holder of Facility

Operating License Nos. DPR–77 and

DPR–79, which authorizes operation of

Sequoyah Nuclear Plant, Units 1 and 2

located in Hamilton County, Tennessee.

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

that use 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

requested licensees to submit plans and

schedules for resolving the Thermo-Lag

issue. The NRC staff has obtained and

reviewed all licensees’ corrective plans

and schedules. The staff is concerned

that some licensees may not be making

adequate progress toward resolving the

plant-specific 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. For plants that have completion

action scheduled beyond 1997,

including Sequoyah Nuclear Plant,

Units 1 and 2, the NRC staff has met

with the licensees to 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.

TVA was one of the licensees with

which the NRC staff held meetings. At

the May 30, 1997, meeting, the NRC

staff reviewed with TVA the schedule of

Thermo-Lag corrective actions for the

Sequoyah units described in the

handout presented to the NRC during

that meeting. Based on the information

provided during the meeting, as well as

a subsequent letter dated June 25, 1997,

the NRC staff has concluded that the

schedules presented by TVA 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, (3)

the need to perform certain plant

modifications during outages as

opposed to those that can be performed

while the plant is at power, and (4)

integration with other significant, but

unrelated issues that TVA is addressing

at its plant. In order to remove

compensatory measures such as fire

watches, it has been determined that

resolution of the Thermo-Lag corrective

actions by TVA must be completed in

accordance with current schedules. By

letter dated April 29, 1998, the NRC

staff notified TVA of its plan to

incorporate TVA’s schedule

commitment into a requirement by

issuance of an order and requested

consent from the Licensee. By letter

dated May 13, 1998, TVA provided its

consent to issuance of a Confirmatory

Order.

III

The Licensee’s commitment as set

forth in its letter of May 13, 1998, is

acceptable and is necessary for the NRC

to conclude that public health and

safety are reasonably assured.

To preclude any schedule slippage

and to assure public health and safety,

the NRC staff has determined that the

Licensee’s commitment in its May 13,

1998, letter be confirmed by this Order.

The Licensee has agreed to this action.

Based on the above, and the Licensee’s

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 Tennessee Valley Authority (TVA)

shall complete final implementation of

Thermo-Lag 330–1 fire barrier corrective

actions at the Sequoyah Nuclear Plant, Units

1 and 2 as described in the TVA submittal

dated June 25, 1997. Walkdowns,

evaluations, and upgrades will be completed

by June 30, 1999.

The Director, Office of Nuclear Reactor

Regulation, may relax or rescind, in

writing, any provisions of this

Confirmatory Order upon a showing by

the Licensee of good cause.

V

Any person adversely affected by this

Confirmatory Order, other than the

Licensee, 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, and 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: Rulemaking and

Adjudications Staff, Washington, D.C.

20555. 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, to the Deputy

Assistant General Counsel for

Enforcement at the same address, to the

Regional Administrator, NRC Region II

at the Atlanta Federal Center, 23 T85, 61

Forsyth Street, SW., Atlanta, Georgia

30303–3415, and to the Licensee. 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 18th day

of June 1998.
Federal Register / Vol. 63, No. 121 /Wednesday, June 24, 1998 / Notices 34491

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director Office of Nuclear Reactor Regulation.

[FR Doc. 98–16745 Filed 6–23–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–424 and 50–425]

Southern Nuclear Operating Company,

Inc., et al.; Vogtle Electric Generating

Plant, Units 1 and 2; Environmental

Assessment and Finding of No

Significant Impact

The U.S. Nuclear Regulatory

Commission (the Commission) is

considering issuance of amendments to

Facility Operating License Nos. NPF–68

and NPF–81 issued to Southern Nuclear

Operating Company, Inc., et al. (the

licensee), for operation of the Vogtle

Electric Generating Plant (VEGP), Units

1 and 2, respectively, located in Burke

County, Georgia.

Environmental Assessment

Identification of the Proposed Action

The proposed action would change

the common VEGP Technical

Specifications to allow an increase in

the Unit 1 spent fuel storage capacity

from 288 to 1476 fuel assemblies. The

increase in spent fuel storage capacity is

achieved by replacing the existing spent

fuel storage racks, a process referred to

herein as ‘‘reracking.’’ The proposed

action is in accordance with the

licensee’s application for license

amendments dated September 4, 1997,

as supplemented by letters dated

November 20, 1997, May 19 and June

12, 1998.

The Need for the Proposed Action

The VEGP spent fuel pools (SFPs) are

operated as a single facility and accept

spent fuel from both Units 1 and 2. The

VEGP Unit 2 spent fuel pool has a

storage capacity of 2098 fuel assemblies.

Under current conditions, the SFPs will

lose the capacity for a full-core off-load

(193 fuel assemblies) in the year 2005.

There are no independent commercial

spent fuel storage facilities operating in

the U.S., nor are there any domestic

reprocessing facilities; therefore, the

projected loss of storage capacity in the

VEGP SFPs would affect the licensee’s

ability to operate VEGP. The proposed

amendments are needed to ensure the

capability of full-core off-load until the

year 2015.

Environmental Impacts of the Proposed

Action

Radiological Impacts

VEGP has waste treatment systems

designed to collect and process waste

that may contain radioactive material.

The radioactive waste treatment systems

were evaluated in the ‘‘Final

Environmental Statement Related to the

Operation of Vogtle Electric Generating

Plant,’’ NUREG–1087, March 1985. The

SFP cooling and purification system is

designed to remove the decay heat

generated by stored spent fuel

assemblies and to clarify and purify the

water to permit unencumbered access to

the plant fuel storage area and maintain

optical clarity of the SFP water.

Liquid Radioactive Waste

It is not expected that there will be a

significant increase in the liquid release

of radionuclides from the plant as a

result of the SFP reracking

modifications. The SFP cooling and

purification system operates as a closed

system. The SFP demineralizer resin

removes soluble radioactive materials

from the SFP water. A small increase in

activity on the filters and demineralizers

may occur during the installation of the

new racks because of the more frequent

fuel shuffling and underwater pressure

washing of the old racks during

removal. However, the amount of

radioactivity released to the

environment as a result of the proposed

reracking is expected to be negligible.

Solid Radioactive Waste

The existing spent fuel racks in the

VEGP Unit 1 SFP will be removed from

the site by a salvage company. After

usable material has been salvaged, the

remainder will be volume reduced and

disposed of at the Barnwell, South

Carolina, facility. In a worst-case

scenario, with no salvageable material

and no volume reduction, the resulting

material would represent 44 percent of

the expected solid waste volume

associated with VEGP Units 1 and 2 for

1998; however, this volume is not

significant when viewed over the 40-

year operational lifetime of the VEGP

facility.

In addition to the spent fuel

assemblies themselves, the only other

solid radioactive waste generated by the

SFP is the SFP polisher resin, which is

used for water clarity. As indicated in

the licensee’s submittal of September 4,

1997, these resins are replaced

approximately once per refueling cycle.

No additional spent resins are expected

to be generated by the pool cleanup

system as a result of the expanded spent

fuel storage capability; therefore, no

significant increase in the volume of

solid radioactive waste associated with

these resins is expected with the

proposed amendments.

Radioactive Material Released to the

Atmosphere

The only radioactive gas of

significance that could be attributable to

storing additional spent fuel assemblies

for a longer period of time, made

possible as a result of the proposed

reracking, would be the noble gas

radionuclide krypton-85 (Kr-85).

Experience has demonstrated that after

spent fuel has decayed 4 to 6 months,

there is no longer a significant release of

fission products, including Kr-85, from

stored spent fuel containing cladding

defects. The licensee has stated that in

the past 2 years, the Kr-85

concentrations measured from the fuel

storage area ventilation release point

have been negligible and the licensee

expects that enlarging the storage

capacity of the SFP will have no effect

on the average annual quantities of Kr85 released to the atmosphere.

Iodine-131 released from spent fuel

assemblies to the SFP water will not be

significantly increased as a result of the

expansion of the fuel storage capacity

since the iodine-131 inventory in the

fuel will decay to negligible levels

between refuelings.

Most of the tritium in the SFP water

results from activation of boron and

lithium in the primary coolant during

power operation. A relatively small

amount of tritium is produced during

reactor operation by the fission process

within the reactor fuel. The subsequent

diffusion of the tritium through the fuel

and cladding represents a small

contribution to the total amount of

tritium in the SFP water. Tritium

releases from the fuel assemblies occur

mainly during reactor operation and, to

a limited extent, shortly after shutdown.

Thus, expanding the SFP capacity will

not increase the tritium concentration in

the SFP.

Most airborne releases of tritium and

iodine from nuclear power plants result

during refuelings from evaporation of

reactor coolant, which contains tritium

and iodine in higher concentrations

than in the SFP. The storage of

additional spent fuel assemblies in the

SFP is not expected to significantly

increase the SFP bulk water

temperature, and, therefore, evaporation

rates from the SFP are not expected to

significantly increase. Consequently, it

is not expected that there will be any

significant change in the annual release

of tritium or iodine as a result of the

proposed modifications from that

previously evaluated in NUREG–1087.