Rochester Gas and Electric Corporation; R.E. Ginna Nuclear Power Plant; Environment Assessment and Finding of No Significant Impact

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 (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 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

[[Page 39297]]

bulk water temperature above the 150 deg.F used in the design analysis

and, therefore, evaporation rates from the SFP are not expected to

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

the FES.

Solid Radioactive Wastes

Spent resins are generated by the spent fuel pool purification

system. These spent resins are replaced every 2 to 3 years and are

disposed of as solid radioactive waste. The licensee will clean the

floor of the SFP using a vacuum system before any work is done and

after each of the old Region I fuel rack modules is removed. The

licensee also plans on vacuuming the old Region I fuel rack modules

before removal from the SFP. The licensee will do this in order to

remove as much of the source term as possible (to minimize personnel

dose), to minimize the generation of spent resins, and to ensure visual

clarity in the SFP to facilitate diving operations and SFP rack change

out. On the basis of experience gained following the 1984-1985 SFP

modification, the licensee concludes that the additional fuel storage

made possible by the increased storage capacity will not result in a

significant change in the generation of solid radwaste (in the form of

spent resins).

Prior to removal from the SFP, the three Region I fuel rack modules

will be vacuumed and hydrolazed to remove any loose crud from the

modules. The fuel rack modules will then be decontaminated to less than

200 mrem/hr and will be either shipped offsite intact or will be cut up

and shipped offsite. If shipped intact, the modules will be dried and

bagged first. Otherwise, the modules will be cut up into small enough

pieces to fit into ``low specific activity'' radwaste boxes. The

licensee has stated that the shipping containers and procedures will

conform to all applicable regulations set forth by the U.S. Department

of Transportation (DOT) as well as the requirements of any State DOT

office through which the shipment may pass and the requirements of the

American Association of State Highway and Transportation Officials.

Liquid Radioactive Wastes

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

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, due to the more frequent fuel shuffling and underwater

hydrolazing 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.

Occupational Dose Consideration

Operating experience has shown that area dose rates in the vicinity

of the SFP are 1.0 to 2.0 mrem/hr, regardless of the quantity of fuel

stored in the SFP. These dose rates may increase slightly during

refueling operations due to crud deposits spalling from spent fuel

assemblies and to activities carried into the pool from the primary

system, resulting in slightly higher concentrations of radionuclides in

the SFP. However, licensee experience to date has not indicated a major

increase in dose rates as a consequence of refueling. The licensee has

calculated the expected dose rates at locations of interest outside the

concrete SFP walls to determine how the increase in fuel capacity will

affect the adjacent area dose rates. The licensee has determined that

the resulting dose rates are well within the Radiation Zone II limits

(2.5 mrem/hr) for all passageways adjacent to the SFP which can be

accessed by personnel.

The total collective occupational dose to plant workers as a result

of the reracking operation is estimated to be between 8 and 12 person-

rem. When the licensee performed an SFP rerack in 1984-1985, the

resulting total collective occupational dose received was 14 person-

rem. The licensee plans on incorporating the lessons learned from this

earlier reracking operation to reduce overall doses during the upcoming

reracking operation. The upcoming reracking operation will follow

detailed procedures prepared with full consideration of as low as is

reasonably achievable (ALARA) principles. On the basis of its review of

the Ginna proposal, the staff concludes that the Ginna SFP rack

modification can be performed in a manner that will ensure that doses

to workers will be maintained ALARA.

Accident Considerations

In its application, the licensee evaluated the possible

consequences of six hypothetical accidents involving fuel in the SFP.

Because the licensee uses single failure proof cranes for the lifting

of heavy loads in the vicinity of the SFP, four of these accidents are

deemed not plausible. The licensee evaluated the other two hypothetical

accidents--the fuel handling accident and the tornado missile accident-

to determine the thyroid and whole-body doses at the Exclusion Area

Boundary, Low Population Zone (LPZ), and Control Room. The proposed

reracking of the Ginna SFP will not affect any of the assumptions or

inputs used in evaluating the dose consequences of either of these

hypothetical accidents.

The NRC staff reviewed the licensee's analysis and performed

confirmatory calculations to check the acceptability of the licensee's

doses. The staff's calculations confirmed that the thyroid doses at the

EAB, LPZ, and Control Room from either a fuel handling accident or a

tornado missile accident meet the acceptance criteria and that the

licensee's calculations are acceptable. The results of the staff's

calculations are presented in the Safety Evaluation to be issued with

the license amendment.

In summary, the proposed action will not increase the probability

or consequences of accidents, no changes are being made to radioactive

waste treatment systems or in the types of any radioactive effluents

that may be released offsite, and the proposed action will not result

in a significant increase in occupational or offsite radiation

exposure. Accordingly, the Commission concludes that there are no

significant radiological environmental impacts associated with the

proposed action.

With regard to potential nonradiological impacts, the proposed

action does not affect nonradiological plant effluents and has no other

nonradiological environmental impact.

Accordingly, the Commission concludes that there are no significant

environmental impacts associated with the proposed action.

Alternatives to the Proposed Action

Since the Commission has concluded there is no significant

environmental impact associated with the proposed action, any

alternatives with equal or greater environmental impact need not be

evaluated. 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 for the R.E.

Ginna Nuclear Power Plant dated December 1973.

[[Page 39298]]

Agencies and Persons Consulted

In accordance with its stated policy, on May 19, 1998, the staff

consulted with Hal Brotie of the New York State Energy Research and

Development Authority, regarding the environmental impact of the

proposed action. The State official had no comments.

Finding of No Significant Impact

Based upon the environmental assessment, the Commission concludes

that the proposed action will not have a significant effect on the

quality of the human environment. Accordingly, the Commission has

determined not to prepare an environmental impact statement for the

proposed action.

For further details with respect to the proposed action, see the

licensee's letter dated March 31, 1997, as supplemented by letters

dated 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, May 8, and May 22, 1998, which 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 Rochester Public Library, 115 South

Avenue, Rochester, New York 14610.

Dated at Rockville, Maryland, this 16th day of July 1998.

For the Nuclear Regulatory Commission.

S. Singh Bajwa,

Director, Project Directorate I-1, Division of Reactor Projects--I/II,

Office of Nuclear Reactor Regulation.

[FR Doc. 98-19541 Filed 7-21-98; 8:45 am]

BILLING CODE 7590-01-P