Federal Register / Vol. 62, No. 183 / Monday, September 22, 1997 / Notices 49539

considered, but off-site doses from these

release points were found to be

negligible.

SMC submitted March 1996

measurement data from stack emissions

showing doses less than 1 millirem

(mrem) per year at the fence line under

nominal conditions. Conservative

estimates of the expected effluent

release rates were calculated by the NRC

staff using assumptions, including the

following: (1) the use of conservative

values for the efficiencies of baghouse

filters based upon the possibility of

undetected filter bag breakages and (2)

a ground-level release point for both

baghouses. The radiation doses resulting

from atmospheric releases were

estimated using the CAP88–PC (Clean

Air Assessment Package 1988) Version

1.0 computer code. The maximally

exposed individual was located at the

fence line, which was 250 meters (820)

feet) south of the SMC facility. The

Total Effective Dose Equivalent (TEDE)

to the nearest resident is estimated to be

less than 9 mrem per year from all

pathways. Inhalation intakes accounted

for greater than 85 percent of the total

radiation dose. Thorium-232 was the

dominant dose contributor, accounting

for about 30 percent of the total dose

This estimated radiation dose is less

than the 100 mrem per year limit

established by NRC in 10 CFR 20.1301

and the 10 millirem per year dose

constraint for air emissions in 10 CFR

20.1101.

The population within 80 km (50

miles) of SMC’s facility is about

6,766,961 people, based on 1994 census

data. The collective dose to the

surrounding population is expected to

be less than 7 person-rem per year.

Based on an average background

radiation dose of about 0.3 rem per year

for individuals in the U.S. from natural

sources, the same population would

receive about 2,00,000 person-rem per

year from background radiation. Thus,

the collective radiation dose associated

with atmospheric releases from the

SMC’s facility is a small percentage of

the collective radiation dose from

natural background radiation for these

same people.

Accident Evaluation

In the EA, NRC staff evaluated one

accident as the bounding accident: the

release of dust from a baghouse or silo.

This accident assumed that 10,000 kg of

dust were released from structural

failure of a baghouse. Calculated release

fractions were 4 to 5×10 ¥3. Other

accidents were determined to be within

the bounds of this accident because both

quantities and form of the material

made larger dispersions unlikely. This

bounding accident was calculated as a

result in an exposure of less than 6

mrem TEDE to the nearest resident. The

expected population dose from this

accident would be no greater than 0.9

person-rem.

Agencies and Persons Consulted

Discussions were held with

representatives from the State of New

Jersey Department of Environmental

Protection and the U.S. Environmental

Protection Agency at various times

throughout the preparation of the EA.

NRC consulted SMC representatives in

preparing this document.

Conclusion

On the basis of this Environmental

Assessment, NRC has concluded that

the environmental impacts from the

proposed action would not be

significant.

Finding of No Significant Impact

The NRC has prepared an EA related

to the renewal of Source Material

License SMB–743. On the basis of the

assessment, the NRC has concluded that

environmental impacts that would be

created by the proposed action would

not be significant and do not warrant

the preparation of an Environmental

Impact Statement. Accordingly, NRC

has determined that a Finding of No

Significant Impact is appropriate.

The EA, the license renewal

application dated September 15, 1995,

and the documents related to this

proposed action are available for public

inspection and copying at the

Commission’s Public Document Room

at the Gelman Building, 2120 L Street

NW, Washington, DC. Anyone with

questions or comments about this

proposed action should contact Ms.

Heather Astwood, NRC’s Project

Manager for the facility, at Mail Stop T–

8D–14, U.S. NRC, Washington, D.C.

20555 or in (301) 415–5819.

Dated at Rockville, Maryland, this 16th day

of September, 1997.

For the Nuclear Regulatory Commission.

Michael F. Weber,

Chief, Licensing Branch, Division of Fuel

Cycle Safety and Safeguards, Office of

Nuclear Material Safety and Safeguards.

[FR Doc. 97–25078 Filed 9–19–97; 8:45 am]

BILLING CODE 7590–01–M

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–321 and 50–366]

Edwin I. Hatch Nuclear 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 an exemption

from certain requirements of its

regulations with respect to Facility

Operating Licenses DPR–57 and NPF–5

issued to Southern Nuclear Operating

Company, Inc., et al. (Southern Nuclear,

or the licensee) for operation of the

Edwin I. Hatch Nuclear Plant, Units 1

and 2, located in Appling County,

Georgia.

Environmental Assessment

Identification of Proposed Action

The proposed action is in accordance

with the licensee’s application dated

July 2, 1997, for exemption from certain

requirements of 10 CFR 73.55,

‘‘Requirements for Physical Protection

of Licensed Activities in Nuclear Power

Reactors Against Radiological

Sabotage.’’ The exemption would allow

photo identification badges to be taken

offsite by individuals not employed by

the licensee who have been granted

unescorted access into protected and

vital areas, in light of the

implementation of a hand geometry

biometrics system to control site access

at Hatch.

The Need for the Proposed Action

Pursuant to 10 CFR 73.55, paragraph

(a), Southern Nuclear shall establish and

maintain an onsite physical protection

system and security organization.

Regulation 10 CFR 73.55(d), ‘‘Access

Requirements,’’ paragraph (1), specifies

that the ‘‘licensee shall control all

points of personnel and vehicle access

into a protected area.’’ Regulation 10

CFR 73.55(d)(5) specifies that, ‘‘A

numbered picture badge identification

system shall be used for all individuals

who are authorized access to protected

areas without escort.’’ Section

73.55(d)(5) also states that an individual

not employed by the licensee (i.e.,

contractors) may be authorized access to

protected areas without escort provided

the individual, ‘‘receives a picture badge

upon entrance into the protected area

which must be returned upon exit from

the protected area....’’ Currently,

unescorted access into protected areas at

the Hatch plant is controlled through

the use of a photograph on a badge/

keycard (hereafter referred to as a

‘‘badge’’), which is stored at the access
49540 Federal Register / Vol. 62, No. 183 / Monday, September 22, 1997 / Notices

point when not in use. The security

officers at each entrance station use the

photograph on the badge to visually

identify the individual requesting

access. The badges for Southern Nuclear

employees and contractor personnel

who have been granted unescorted

access are given to the individuals at the

entrance location upon entry and are

returned upon exit. In accordance with

10 CFR 73.55(d)(5), the badges are not

allowed to be taken offsite.

The licensee proposes to implement

an alternate unescorted access control

system that would eliminate the need to

issue and retrieve badges at the entry

point and would allow all individuals

with unescorted access to keep their

badges when departing the site. An

exemption from 10 CFR 73.55(d)(5) is

required to permit contractors to take

their badges offsite instead of returning

them when exiting the site.

Environmental Impacts of the Proposed

Action

Because the proposed action involves

administrative matters within the

protected area as defined in 10 CFR Part

20, the Commission concludes that this

proposed action would result in no

significant radiological impacts. With

regard to potential nonradiological

impacts, the proposed action does not

affect nonradiological plant effluents

and has no other environmental impact.

Therefore, the Commission concludes

that there are no significant

environmental impacts associated with

the proposed action.

Alternative to the Proposed Action

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.

Alternative Use of Resources

This action does not involve the use

of any resources not previously

considered in the Final Environmental

Statement related to operation of the

Edwin I. Hatch Nuclear Plant, Unit 1

dated October 1972, and Unit 2 dated

March 1978.

Agencies and Persons Consulted

In accordance with its stated policy,

on August 22, 1997, the staff consulted

with the Georgia State official, Mr.

James Setser of the Environmental

Protection Division, Georgia Department

of Natural Resources, regarding the

environmental impact of the proposed

action. The State official had no

comments.

Finding of No Significant Impact

Based upon the foregoing

environmental assessment, the

Commission has determined not to

prepare an environmental impact

statement for the proposed exemption.

Accordingly, the Commission has

concluded that the proposed action will

not have a significant effect on the

quality of the human environment.

For further details with respect to this

action, see the request for exemption

dated July 2, 1997, 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

Burke County Public Library, 412

Fourth Street, Waynesboro, Georgia.

Dated at Rockville, Maryland, this 15th day

of September 1997.

For the Nuclear Regulatory Commission.

Herbert N. Berkow,

Director, Project Directorate II–2, Division of

Reactor Projects—I/II, Office of Nuclear

Reactor Regulation.

[FR Doc. 97–25081 Filed 9–19–97; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–338 AND 50–339]

Virginia Electric and Power Company;

North Anna Power Station, Units 1 and

2 Environmental Assessment and

Finding of No Significant Impact

The U.S. Nuclear Regulatory

Commission (the Commission) is

considering issuance of an exemption

from the provisions of 10 CFR 70.24(a)

to Virginia Electric and Power Company

(the licensee) for North Anna Power

Station, Units 1 and 2 (NPS1&2), located

in Louisa County, Virginia.

Environmental Assessment

Identification of Proposed Action

The proposed action would exempt

the licensee from the requirements of 10

CFR 70.24(a), which require a

monitoring system that will energize

clear audible alarms if accidental

criticality occurs in each area in which

special nuclear material (SNM) is

handled, used, or stored. The proposed

action would also exempt the licensee

from the requirements to maintain

emergency procedures for each area in

which this licensed SNM is handled,

used, or stored to ensure that all

personnel withdraw to an area of safety

upon sounding of the alarm, to

familiarize personnel with the

evacuation plan, and to designate

responsible individuals for determining

the cause of the alarm, and to place

radiation survey instruments in

accessible locations for use in such an

emergency.

The proposed action is in accordance

with the licensee’s application for

exemption dated January 28, 1997, as

supplemented March 24, 1997.

The Need for the Proposed Action

The purpose of 10 CFR 70.24(a) is to

ensure that if a criticality were to occur

during the handling, use, or storing of

SNM, personnel would be alerted to that

fact and would take appropriate action.

At a commercial nuclear power plant,

the inadvertent criticality with which 10

CFR 70.24 is concerned could occur

during fuel handling operations. The

SNM that could be assembled into a

critical mass is in the form of nuclear

fuel. The quantity of other forms of

special nuclear materials that is stored

onsite is small enough to preclude

achieving critical mass. Since the fuel is

not enriched beyond 4.3 weight percent

Uranium-235 and commercial nuclear

power plant licensees have procedures

and features that are designed to prevent

inadvertent criticality, the staff has

determined that inadvertent criticality is

not likely to occur during the handling

of the special nuclear material. The

requirements of 10 CFR 70.24(a),

therefore, are not necessary to ensure

the safety of personnel during the

handling of special nuclear materials at

commercial power plants.

Environmental Impacts of the Proposed

Action

The Commission has completed its

evaluation of the proposed action and

concludes that there is no significant

environmental impact if the exemption

is granted. Inadvertent or accidental

criticality will be precluded through the

design of the fuel racks providing

geometric spacing of fuel assemblies in

their storage locations, compliance with

the NPS Technical Specifications (TS),

and administrative controls imposed on

fuel handling procedures.

Appendix A of 10 CFR Part 50,

‘‘General Design Criteria for Nuclear

Power Plants,’’ Criterion 62, requires

that criticality in the fuel storage and

handling system shall be prevented by

physical systems or processes,

preferably by use of geometrically safe

configurations. This is met at NPS1&2,

as identified in section 5.6 of the TS.

Section 5.6.1.1 of the TS states the

geometrically safe configurations for