Federal Register / Vol. 63, No. 140 /Wednesday, July 22, 1998 / Notices 39295

State Archaeological Research Center,

Rapid City, SD.

A detailed assessment of the human

remains was made by South Dakota

State Archaeological Research Center

and Office of the State Archeologist of

Iowa professional staff in consultation

with representatives of the Three

Affiliated Tribes of the Fort Berthold

Reservation.

In 1934, human remains representing

one individual likely to have been

recovered from the Evart’s Village site

(39WW204), Walworth County, SD

during Works Project Administration

road construction. No known individual

was identified. The two associated

funerary objects consist of a white glass

pony bead and a rifle bullet, identified

as possibly a .44-60 calibre Peabody,

Remington, or Sharps.

In 1990, these human remains were

found in the collections of the Conger

House Museum in Washington, IA and

transferred to the Office of the State

Archeologist of Iowa. Museum

documentation suggests these remains

were recovered from the Evart’s Village

site on Fred Brazel’s land near Evart, SD

and given to his brother-in-law, Thomas

Royster of Muscatine, IA. Mr. Royster

may have donated these remains to the

Conger House Museum, as Washington,

IA is near Muscatine. In 1952, an

interview with Mrs. Fred Brazel

revealed that these human remains were

possibly interred as a primary flexed or

secondary bundle burial, placed face up

on top of a layer of cut willow twigs.

Based on skeletal morphology and

associated funerary objects, these

individuals have been determined to be

Native American. Based on the

associated funerary objects, manner of

interment, and geographical location,

the Evart’s Village site has been

identified as a post–1770 Arikara or

Mandan village. Consultation with

representatives of the Three Affiliated

Tribes indicates there were Arikara and

Mandan villages in this area of South

Dakota during the post contact period.

Based on the above mentioned

information, officials of the South

Dakota State Archaeological Research

Center have determined that, pursuant

to 43 CFR 10.2 (d)(1), the human

remains listed above represent the

physical remains of one individual of

Native American ancestry. Officials of

the South Dakota State Archaeological

Research Center have also determined

that, pursuant to 43 CFR 10.2 (d)(2), the

two objects listed above are reasonably

believed to have been placed with or

near individual human remains at the

time of death or later as part of the death

rite or ceremony. Lastly, officials of the

South Dakota State Archaeological

Research Center have determined that,

pursuant to 43 CFR 10.2 (e), there is a

relationship of shared group identity

which can be reasonably traced between

these Native American human remains

and associated funerary objects and the

Three Affiliated Tribes of the Fort

Berthold Reservation.

This notice has been sent to officials

of the Three Affiliated Tribes of the Fort

Berthold Reservation. Representatives of

any other Indian tribe that believes itself

to be culturally affiliated with these

human remains and associated funerary

objects should contact Renee Boen,

Curator, South Dakota State

Archaeological Reserch Center, P.O. Box

1257, Rapid City, SD 57709-1257;

telephone: (605) 394–1936, before

August 22, 1999. Repatriation of the

human remains and associated funerary

objects to the Three Affiliated Tribes of

the Fort Berthold Reservation may begin

after that date if no additional claimants

come forward.

Dated: July 6, 1998.

Francis P. McManamon,

Departmental Consulting Archeologist,

Manager, Archeology and Ethnography

Program.

[FR Doc. 98–19538 Filed 7–21–98; 8:45 am]

BILLING CODE 4310–70–F

NUCLEAR REGULATORY

COMMISSION

[Docket Nos. 50–280 and 50–281]

In the Matter of Virginia Electric and

Power Company; Surry Power Station,

Units 1 and 2; Exemption

I

The Virginia Electric and Power

Company (VEPCO, the licensee) is the

holder of Facility Operating License No.

DPR–32 and Facility Operating License

No. DPR–37, which authorize operation

of the Surry Power Station, Units 1 and

2. The licenses provide that the licensee

is subject to all rules, regulations, and

orders of the U.S. Nuclear Regulatory

Commission (NRC or the Commission)

now or hereafter in effect.

The facility consists of two

pressurized-water reactors at the

licensee’s site located in Surry County,

Virginia.

II

Title 10 of the Code of Federal

Regulations (10 CFR) Section 70.24,

‘‘Criticality Accident Requirements,’’

requires that each licensee authorized to

possess special nuclear material shall

maintain a criticality accident

monitoring system in each area in

which such material is handled, used,

or stored. Sections 70.24 (a)(1) and (a)(2)

specify detection and sensitivity

requirements that these monitors must

meet. Section 70.24(a)(1) also specifies

that all areas subject to criticality

accident monitoring must be covered by

two detectors. Section 70.24(a)(3)

requires licensees to maintain

emergency procedures for each area in

which this licensed special nuclear

material is handled, used, or stored, and

provides (1) that the procedures ensure

that all personnel withdraw to an area

of safety upon the sounding of a

criticality accident monitor alarm, (2)

that the procedures must include drills

to familiarize personnel with the

evacuation plan, and (3) that 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.

Section 70.24(b)(1) requires licensees to

have a means by which to quickly

identify personnel who have received a

dose of 10 rads or more. Section

70.24(b)(2) 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.

Section 70.24(c) exempts Part 50

licensees from the requirements of 10

CFR 70.24(c) for special nuclear

material used or to be used in the

reactor. Subsection 70.24(d) 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

On August 21, 1997, the NRC granted

an exemption from the requirements of

10 CFR 70.24 reflecting the licensee’s

use of fuel enriched to 4.1 weight

percent U235. By letter dated January

14, 1998, VEPCO requested a revised

exemption from 10 CFR 70.24(a) based

on the use of fuel enriched to 4.3 weight

percent U235. The Commission has

reviewed the licensee’s submittal and

has determined that inadvertent

criticality is not likely to occur in

special nuclear materials handling or

storage areas at Surry Power Station,

Units 1 and 2. The quantity of special

nuclear material other than fuel that is

stored on site is small enough to

preclude achieving a critical mass.

The purpose of the criticality

monitors required by 10 CFR 70.24 is to
39296 Federal Register / Vol. 63, No. 140 /Wednesday, July 22, 1998 / Notices

ensure that if a criticality were to occur

during the handling of special nuclear

material, personnel would be alerted to

that fact and would take appropriate

action. Although the staff has

determined that such an accident is not

likely to occur, the licensee has

radiation monitors, as required by

General Design Criteria 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(a).

IV

The Commission has determined that,

pursuant to 10 CFR 70.14, this

exemption as revised 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 following revised exemption:

The Virginia Electric and Power

Company is exempt from the

requirements of 10 CFR 70.24(a) for the

Surry Nuclear Power Station, Unit 1 and

Unit 2.

Pursuant to 10 CFR 51.32, the

Commission has determined that the

granting of this revised exemption will

have no significant impact on the

quality of the human environment (63

FR 38196).

This revised exemption is effective

upon issuance.

Dated at Rockville, Maryland this 15th day

of July 1998.

For the Nuclear Regulatory Commission.

Samuel J. Collins,

Director, Office of Nuclear Reactor

Regulation.

[FR Doc. 98–19539 Filed 7–21–98; 8:45 am]

BILLING CODE 7590–01–P

NUCLEAR REGULATORY

COMMISSION

[Docket No. 50–244]

Rochester Gas and Electric

Corporation; R.E. Ginna Nuclear Power

Plant; Environment Assessment and

Finding of No Significant Impact

The U.S. Nuclear Regulatory

Commission (the Commission) is

considering issuance of an amendment

to Facility Operating License No. DRP–

18, issued to Rochester Gas and Electric

Corporation (the licensee), for operation

of the R.E. Ginna Nuclear Power Plant,

located in Wayne County, New, York.

Identification of the Proposed Action

The proposed action would modify

the spent fuel pool (SFP) by replacing

the three Region 1 rack modules with

seven new borated stainless steel rack

modules scheduled for implementation

in 1998. Six new peripheral modules

would be added at some future date.

Two of the seven new modules planned

to be installed in 1998 would be

designated as part of Region 2,

effectively increasing the Region 2 area.

The other five new modules would

compose Region 1, resulting in a total of

294 storage positions in Region 1.

Region 2, with 1075 storage positions,

would consist of three rack types, Type

1, Type 2, and Type 4. Type 1 cells are

the Boraflex cells that form Region 2 for

the existing license. Two racks of Type

2 cells, containing borated stainless

steel (BSS) absorber plates, would be

added to increase the storage capacity of

Region 2. In addition, the capacity of

Region 2 could be increased in the

future by the addition of Type 4 racks,

which also contain BSS absorber plates.

The amendment would also increase the

boron concentration from 300 ppm to

2300 ppm.

The proposed action is in accordance

with the licensee’s application for

amendment dated March 31, 1997, as

supplemented June 18, 1997, October

10, 1997, November 11, 1997, December

22, 1997, January 15, 1998, January 27,

1998, March 20, 1998, April 23, 1998,

April 27, 1998, and May 8, 1998.

The Need for the Proposed Action

The proposed action would modify

the spent fuel pool to accommodate

storage of spent fuel until the expiration

of the Ginna Station license in 2009.

The current configuration of the Ginna

spent fuel storage pool consists of two

regions. Region 1 consists of stainless

steel racks with 176 storage locations in

a checker board pattern. Region 2

consists of stainless steel racks with

boraflex and with 840 storage locations.

This provides a total of 1016 storage

locations. The proposed amendment

would replace the Region 1 racks with

borated stainless steel racks. Two

locations are proposed in Region 1, one

with borated stainless steel that would

accommodate 187 storage locations and

one with borated stainless steel in a

checker board pattern that would

accommodate 292 storage locations.

This would provide a total of 1319

storage locations which would provide

enough storage locations for storage of

spent fuel beyond the expiration of the

license in 2009.

Environmental Impacts of the Proposed

Action

Radioactive Waste Treatment

The Ginna Nuclear Power Plant uses

waste treatment systems designed to

collect and process gaseous, liquid, and

solid waste that might contain

radioactive material. These radioactive

waste treatment systems are evaluated

in the Final Environmental Statement

(FES) dated December 1973. The

proposed rerack will not involve any

change in the waste treatment systems

described in the FES.

Gaseous Radioactive Wastes

The only radioactive gas of

significance that could be attributable to

storing additional spent fuel assemblies

for a longer period of time would be the

noble gas radionuclide Krypton-85 (Kr85). 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 the Kr-85 noble gases are not

normally released from the Auxiliary

Building on a continuous basis and

enlarging the storage capacity of the SFP

will have no effect on the average

annual quantities of Kr-85 released to

the atmosphere.

Iodine-131 released from spent fuel

assemblies to the SFP water will not be

significantly increased due to the

expansion of the fuel storage capacity

since the Iodine-131 inventory in the

fuel will decay to negligible levels

between refuelings.

The amount of tritium in the SFP

water will not be affected by the

proposed changes. Most of the tritium in

the SFP water results from activation of

boron and lithium in the primary

coolant. 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 activity 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 increase the SFP