[Federal Register Volume 63, Number 140 (Wednesday, July 22, 1998)]
[Notices]
[Pages 39296-39298]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-19541]


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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