29762 Federal Register / Vol. 63, No. 104 / Monday, June 1, 1998 / Notices

requirements of the Atomic Energy Act

of 1954, as amended (the Act), and the

Commission’s rules and regulations.

The Commission has made appropriate

findings as required by the Act and the

Commission’s rules and regulations in

10 CFR Chapter I, which are set forth in

the license amendment.

Notice of Consideration of Issuance of

Amendment to Facility Operating

License and Opportunity for a Hearing

in connection with this action as it

applies to the Improved Technical

Specifications was published in the

Federal Register on July 22, 1997 (62 FR

39283). No request for a hearing or

petition for leave to intervene was filed

following this notice. The Commission

has prepared an Environmental

Assessment related to the action and has

determined not to prepare an

environmental impact statement. Based

upon the environmental assessment, the

Commission has concluded that the

issuance of the amendment will not

have a significant effect on the quality

of the human environment (63 FR

13078, dated March 17, 1998).

Notices of Consideration of Issuance

of Amendment to Facility Operation

License, Proposed No Significant

Hazards Consideration Determination,

and Opportunity for a Hearing in

connection with this action as it applies

to the revised combinations of

emergency core cooling systems/

subsystems that may be out of service

and to the relaxed required flowrates for

the core spray, the low pressure coolant

injection, and the high pressure coolant

injection systems were published in the

Federal Register on December 31, 1997

(62 FR 68306) and February 11, 1998 (63

FR 6986), respectively. No request for a

hearing or petition for leave to intervene

was filed following these notices and no

significant hazards consideration

comments were received.

For further details with respect to the

action see (1) the application for

amendment dated October 30, 1996, as

supplemented by letters dated June 10,

September 5, 17, and 30, October 16,

November 18 and 21, December 8 and

15, 1997, January 2, 5, 12, 22 and 23,

February 10, 26, March 23, 31, and

April 17, 1998, (2) Amendment No. 223

to License No. DPR–49, (3) the

Commission’s related Safety Evaluation,

and (4) the Commission’s

Environmental Assessment. All of these

items 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 local public document room located

at the Cedar Rapids Public Library, 500

First Street, SE., Cedar Rapids, IA

52401.

Dated at Rockville, Maryland, this 22nd

day of May 1998.

For the Nuclear Regulatory Commission.

Richard J. Laufer,

Project Manager, Project Directorate III–3,

Division of Reactor Projects, Office of Nuclear

Reactor Regulation.

[FR Doc. 98–14392 Filed 5–29–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–298]

In the Matter of: Nebraska Public

Power District (Cooper Nuclear

Station); Exemption

I

The Nebraska Public Power District

(the licensee) is the holder of Facility

Operating License No. DRP–46, which

authorizes operation of the Cooper

Nuclear Station. The license provides,

among other things, that the licensee is

subject to all rules, regulations, and

orders of the Commission now or

hereafter in effect.

The facility consists of one boilingwater reactor at the licensee’s site

located in Nemaha County, Nebraska.

II

Section 70.24 of Title 10 of the Code

of Federal Regulations, ‘‘Criticality

Accident Requirements,’’ requires that

each licensee authorized to possess

special nuclear material (SNM) shall

maintain a criticality accident

monitoring system in each area where

such material is handled, used, or

stored. Subsections (a)(1) and (a)(2) of

10 CFR 70.24 specify detection and

sensitivity requirements that these

monitors must meet. Subsection (a)(1)

also specifies that all areas subject to

criticality accident monitoring must be

covered by two detectors. Subsection

(a)(3) of 10 CFR 70.24 requires licensees

to maintain emergency procedures for

each area in which this licensed SNM

is handled, used, or stored and provides

that (1) the procedures ensure that all

personnel withdraw to an area of safety

upon the sounding of a criticality

accident monitor alarm, (2) the

procedures must include drills to

familiarize personnel with the

evacuation plan, and (3) the procedures

designate responsible individuals for

determining the cause of the alarm and

placement of radiation survey

instruments in accessible locations for

use in such an emergency. Subsection

(b)(1) of 10 CFR 70.24 requires licensees

to have a means to identify quickly

personnel who have received a dose of

10 rads or more. Subsection (b)(2) of 10

CFR 70.24 requires licensees to

maintain personnel decontamination

facilities, to maintain arrangements for a

physician and other medical personnel

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 Cooper Nuclear

Station 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 Cooper Nuclear

Station, 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 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
Federal Register / Vol. 63, No. 104 / Monday, June 1, 1998 / Notices 29763

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 February 23, 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 submittals and has

determined that Cooper Nuclear Station

meets the applicable criteria. Criteria 2

and 3 are not applicable to the Cooper

Nuclear Station since the fresh fuel

storage racks are not currently in use

and administrative controls prevent

their use. Therefore, the staff has

determined that it is extremely unlikely

for an inadvertent criticality to occur in

SNM handling or storage areas at

Cooper Nuclear Station.

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;

furthermore, the licensee has radiation

monitors, as required by General Design

Criterion 63, in fuel storage and

handling areas. These monitors will

alert personnel to excessive radiation

levels and allow them to initiate

appropriate safety actions. The low

probability of an inadvertent criticality,

together with the licensee’s adherence

to General Design Criterion 63,

constitute good cause for granting an

exemption to the requirements of 10

CFR 70.24.

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 the Nebraska

Public Power District an exemption

from the requirements of 10 CFR 70.24.

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 28012).

This exemption is effective upon

issuance.

Dated at Rockville, Maryland, this 22nd

day of May 1998.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 98–14387 Filed 5–29–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–259, 50–260 and 50–296]

Tennessee Valley Authority; Notice of

Consideration of Issuance of

Amendment to Facility Operating

Licenses and Opportunity for a

Hearing

The U.S. Nuclear Regulatory

Commission (NRC, the Commission) is

considering issuance of an amendment

to Facility Operating License Nos. DPR–

33, DPR–52 and DPR–68 issued to the

Tennessee Valley Authority (TVA or the

licensee) for operation of the Browns

Ferry Nuclear Plant (Browns Ferry,

BFN), Units 1, 2 and 3, located in

Limestone County, Alabama.

Originally, in a letter dated September

6, 1996, the licensee proposed changes

for a full conversion from the current

Technical Specifications (TS) to a set of

TS based on NUREG–1433, Revision 1,

‘‘Standard Technical Specifications for

General Electric Plants, BWR/4,’’ dated

April 1995. NUREG–1433 has been

developed through working groups

composed of both NRC staff members

and the BWR/4 owners and has been

endorsed by the staff as part of an

industry-wide initiative to standardize

and improve TS. As part of this

submittal, the licensee applied the

criteria contained in the Commission’s

‘‘Final Policy Statement on Technical

Specification Improvements for Nuclear

Power Reactors (Final Policy

Statement),’’ published in the Federal

Register on July 22, 1993 (58 FR 39132),

to the current Browns Ferry TS, and,

using NUREG–1433 as a basis,

developed a proposed set of improved

TS for BFN. The criteria in the final

policy statement were subsequently

added to 10 CFR 50.36, ‘‘Technical

Specifications,’’ in a rule change which

was published in the Federal Register

(FR) on July 19, 1996 (60 FR 36953) and

became effective on August 18, 1995. In

addition to the above changes related to

conversion of the current TS to be

similar to the Improved Standard

Technical Specifications (ISTS) in

NUREG 1433, the licensee proposed

three less restrictive changes that are not

considered within the scope of the

normal ISTS conversion process. These

proposed additional changes would (1)

allow two Residual Heat Removal (RHR)

Low Pressure Coolant Injection (LPCI)

pumps (two in one loop or one in both

loops) to be inoperable for 7 days

provided other low pressure emergency

core cooling system (ECCS) pumps are

operable. Current TS requirements

allow only one LPCI pump to be

inoperable, and (2) require only two

ECCS subsystems to be operable during

shutdown. The current TS, which

define subsystems in the same manner

as the ISTS, require three subsystems to

be operable, and (3) reduce the number

of RHR Service Water pumps required

to be operable under certain conditions.

The licensee’s proposed changes in its

application dated September 6, 1996,

including the three additional changes,

were originally noticed on October 23,

1996 (61 FR 55026).

By letters dated June 6, and December

11, 1996, April 11, May 1, August 14,

October 15, November 5 and 14,

December 3, 4, 15, 22, 23, 29, and 30,

1997, January 23, March 12 and 13,

April 16, 20, and 28, May 7, 14, and 19,

1998, the licensee provided

supplemental information, and

proposed additional changes. Some of

these additional changes were ‘‘less

restrictive and plant specific changes’’

that were not included in the original

notice. They are addressed in this

notice. Other changes are related to

conversion of the current TS to those

similar to the ISTS in NUREG 1433 and

are considered to be within the scope of

original FR notice dated October 23,

1996, and therefore, are not addressed

in this notice.

The additional ‘‘less restrictive and

plant specific changes’’ involve: (1)

plant-specific application of generically

approved methodology supporting

extended instrument surveillance

intervals and allowed outage times, (2)

BFN’s operating practice to treat

secondary containment as a single zone

rather than three independent zones for

containment isolation, (3) TS changes to

support installation of a Power Range

Neutron Monitoring System, Average

Power Range Monitor and Rod Block

Monitor TS improvements, and the

Maximum Extended Load Line Limit

analysis, (4) revising the current TS

2.02, consistent with ISTS, to specify

that reactor vessel water level should be

greater than the top of the active

irradiated fuel, instead of specifying

actual water level, (5) proposing an ISTS

to reflect plant-specific design condition

that excludes average U–235 enrichment

of 4.5 weight percent, and (6) TS

changes to allow spiral offload

procedures and adopt a revision to

surveillance requirement 3.3.1.2.4 Note