[Federal Register Volume 65, Number 177 (Tuesday, September 12, 2000)]
[Notices]
[Pages 55059-55061]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 00-23357]
[[Page 55059]]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-302]
Florida Power Corporation; Crystal River Unit 3; Environmental
Assessment and Finding of No Significant Impact
The U.S. Nuclear Regulatory Commission (NRC) is considering
issuance of an amendment to Facility Operating License No. DPR-72
issued to Florida Power Corporation (FPC or the licensee), for
operation of Crystal River Unit 3 (CR-3) located in Citrus County,
Florida.
Environmental Assessment
Identification of the Proposed Action
The proposed action would increase the number of fuel assemblies
that can be stored in the CR-3 spent fuel pools (SFPs) from 1357 fuel
assemblies to 1474 fuel assemblies, an increase of approximately 8
percent, and change the configuration of fresh fuel storage in spent
fuel pool A. In addition, the new spent fuel storage racks will use
Boral as the neutron absorber material, replacing the present neutron
absorber material, Boraflex, which is continuing to degrade.
The proposed action is in accordance with the licensee's
application for amendment dated September 16, 1999, as supplemented by
letters dated May 3 and June 29, 2000.
The Need for the Proposed Action
The currently available storage capacity for spent fuel at CR-3,
allowing for the required reserve capacity to accommodate a full core
offload, is projected to be exceeded in the year 2013. The CR-3
operating license has an expiration date of December 3, 2016. Thus, the
additional 117 locations for storage of fuel assemblies are necessary
to provide adequate spent fuel storage capacity for the remainder of
the CR-3 operating license. In addition, the existing racks utilize
Boraflex as the neutron absorber material. The new spent fuel storage
racks utilize Boral as the neutron absorber material, which will
minimize the water clarity problems associated with use of Boraflex.
Environmental Impacts of the Proposed Action
Radioactive Waste Treatment
CR-3 uses waste treatment systems designed to collect and process
gaseous, liquid, and solid waste that might contain radioactive
material. These radioactive waste treatment systems were evaluated in
the Final Environmental Statement (FES) dated May 1973. The proposed
changes to the SFP will not involve any change in the waste treatment
systems described in the FES.
Gaseous Radioactive Wastes
The storage of additional spent fuel assemblies in the pool is not
expected to affect the releases of radioactive gases from the spent
fuel pool. Gaseous fission products such as Krypton-85 and Iodine-131
are produced by the fuel in the core during reactor operation. A small
percentage of these fission gases can be released to the reactor
coolant from the small number of fuel assemblies that are expected to
develop leaks during reactor operation. During refueling operations,
some of these fission products would then enter the pools and be
subsequently released into the air. At CR-3, there has been no measured
Krypton-85 release from the fuel building ventilation system for the 2
years preceding the September 16, 1999, submittal. Since the frequency
of refueling (and, therefore, the number of freshly offloaded spent
fuel assemblies stored in the pools at any one time) will not increase,
there will be no increase in the amounts of these types of fission
products released to the atmosphere as a result of the increased pool
fuel storage capacity.
The increased heat load on the pool from the storage of additional
spent fuel assemblies was determined by the licensee to be
insignificant, and therefore there would be no significant increase in
the pools' evaporation rate. Therefore, no increase in the amount of
gaseous tritium released from the pool is expected. The overall release
of radioactive gases from CR-3 will remain a small fraction of the
limits of 10 CFR 20.1301.
Solid Radioactive Wastes
Spent resins are generated by the processing of SFP water through
the pools' purification system. These spent resins are disposed of as
solid radioactive waste. Resin replacement is determined primarily by
the requirement for water clarity and is normally done approximately
once per year. No significant increase in the volume of solid
radioactive waste is expected with the expanded storage capacity.
During reracking operations, small amounts of additional waste resin
may be generated by the pools' cleanup systems on a one-time basis.
Additional solid radwaste will consist of the old spent fuel rack
modules themselves, as well as any interferences of pool hardware that
may have to be removed from the pool to permit installation of the new
rack modules. The old racks will be washed down in preparation for
packaging and shipment. Shipping containers and procedures will conform
to Federal regulations as specified in 10 CFR Part 71, ``Packaging and
Transportation of Radioactive Material,'' and to the requirements of
any state through which the shipment may pass, as set forth by the
state's department of transportation.
Liquid Radioactive Wastes
The release of radioactive liquids will not be affected directly as
a result of the SFP modifications. The SFP ion exchanger resins remove
soluble radioactive materials from the pool water. When the resins are
replaced, the small amount of resin sluice water that is released is
processed by the radwaste systems. As previously stated, the frequency
of resin replacement may increase slightly during the installation of
the new racks. However, the increase in the amount of radioactive
liquid released to the environment as a result of the proposed SFP
expansion is expected to be negligible.
Occupational Dose Consideration
Radiation protection personnel at CR-3 will monitor the doses to
the workers during the SFP expansion operations. The total occupational
dose to plant workers as a result of the SFP reracking operations is
estimated to be approximately 3 person-rem, which includes estimates of
person-rem exposures associated with washdown and preparation of the
existing racks for shipping. No diving operations are planned for the
actual rack replacement operation. The dose estimate is comparable to
doses for similar SFP modifications performed at other nuclear plants.
The SFP rack installations will follow detailed procedures prepared
with full consideration of as low as reasonably achievable (ALARA)
principles.
On the basis of its review of the licensee's proposal, the NRC
staff concludes that the CR-3 SFP reracking operations can be performed
in a manner that will ensure that doses to workers will be maintained
ALARA. The estimated dose of 3 person-rem to perform the proposed SFP
reracking operations is a small fraction of the annual collective dose
accrued at CR-3.
Accident Considerations
A fuel handling accident outside the reactor building at CR-3 is
postulated as the dropping of a fuel assembly into the SFP, resulting
in damage to all 208 fuel pins in the dropped fuel assembly. The
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radiological consequences of this accident are based solely on the
damage to the dropped assembly. The replacement racks only increase the
storage capacity of the SFP and do not change the frequency or method
for handling fuel assemblies. The revised fuel storage configuration
does not affect the construction or fuel enrichment of individual fuel
assemblies. Therefore, the probability or consequences of a fuel
handling accident is not increased.
The licensee evaluated spent fuel drop accidents onto the spent
fuel racks, assuming three different orientations, and the dropping of
a rack onto the spent fuel pool floor. The three orientations for the
fuel assembly drops were; (1) Drop of a fuel assembly onto the top of a
rack with the assembly in a vertical position, (2) drop of a fuel
assembly onto the top of a rack with the assembly in an inclined
position, and (3) drop of a fuel assembly through an empty rack cell to
the bottom of the rack. In each case, the rack structure retained the
functional capability to maintain the fuel in a non-critical state. For
orientation 3, the drop to the bottom of the empty rack cell did not
result in penetration of the pool liner.
An analysis was performed to determine the consequences of a rack
drop into SFP B (racks will not be moved over SFP A). The heaviest load
to be lifted as part of the rack replacement project is a rack
currently in SFP B with a weight of 17,715 pounds. The combined weight
of this rack and the lifting rig is less than 20,000 pounds. The load
drop analysis was performed using a bounding load of 20,000 pounds,
assumed to be dropped from the highest lift point of 6 inches above the
spent fuel pool operating deck to the pool floor.
The results of dropping a rack directly onto the SFP floor were the
puncturing of the SFP liner and penetrating the 5-foot thick concrete
floor slab below the liner to a depth of less than 6 inches. The seams
between all sections of concrete are sealed and a waterproof sealant
applied to the inside surfaces of the concrete. The floor and walls of
the CR-3 SFP have a system of leak chases at the welded joints of the
stainless steel liner panels. The leak chase trenches collect liner
leakage and drain by gravity to a leak test hopper/funnel. Isolation
valves are provided in each drain line from the leak chase trenches to
the hopper. These valves will be maintained closed during rack
movements, thereby precluding excessive leakage that might occur
following a load drop. The only non-isolable leakage from the SFP would
be a slow migration of the water from the site of the puncture. The
rate of this leakage would be limited by the low permeability of the
concrete to a negligible value.
CR-3 has various sources of make-up to the SFP. The sources are the
Decay Heat System, the Demineralized Water Supply System, and temporary
fire hoses. Based on the isolation valves being maintained closed, the
negligible leakage rate through the concrete, and the various sources
of make-up, the make-up capability exceeds any leakage resulting from a
rack drop. Uncovery of the fuel stored in the SFP B is precluded, and,
therefore, there is no increase in consequences as a result of a rack
drop onto the SFP floor.
The change in fresh fuel storage configuration in SFP A will result
in the effective neutron multiplication factor remaining well below
0.95. Therefore, there is no reduction in the margin to criticality as
a result of the change in fresh fuel storage configuration in SFP A,
and no increase in the probability of an inadvertent criticality.
The Commission has completed its evaluation of the proposed action
and concludes that the proposed action will not increase the
probability or consequences of accidents, no changes are being made in
the amount or types of any effluents that may be released off site, and
there is no significant increase in occupational or public radiation
exposure. Therefore, there are no significant radiological
environmental impacts associated with the proposed action.
With regard to potential non-radiological impacts, the proposed
action does not involve any historical sites. It does not affect non-
radiological plant effluents and has no other environmental impact.
Therefore, there are no significant non-radiological environmental
impacts associated with the proposed action.
Accordingly, the Commission concludes that there are no significant
environmental impacts associated with the proposed action.
Alternatives to the Proposed Action
Shipping Fuel to a Permanent Federal Fuel Storage/Disposal Facility
Shipment of spent fuel to a high-level radioactive storage facility
is an alternative to increasing the onsite spent fuel storage capacity.
However, the U.S. Department of Energy's (DOE's) high-level radioactive
waste repository is not expected to begin receiving spent fuel until
approximately 2010, at the earliest. To date, no location has been
identified and an interim federal storage facility has yet to be
identified in advance of a decision on a permanent repository.
Therefore, shipping the spent fuel to the DOE repository is not
considered an alternative to increased onsite fuel storage capacity at
this time.
Shipping Fuel to a Reprocessing Facility
Reprocessing of spent fuel from CR-3 is not a viable alternative
since there are no operating commercial reprocessing facilities in the
United States. Therefore, spent fuel would have to be shipped to an
overseas facility for reprocessing. However, this approach has never
been used and it would require approval by the Department of State as
well as other entities. Therefore, shipping fuel to a reprocessing
facility is not a viable option.
Reduction of Spent Fuel Generation
Operation at a reduced power level would decrease the amount of
fuel being stored in the pool and thus increase the amount of time
before full core off-load capacity is lost. However, operating the
plant at a reduced power level would not make effective use of
available resources, and the generation of replacement power would also
result in environmental impacts. Therefore, reducing the amount of
spent fuel generated by reducing power would not result in a
significant improvement in environmental impacts and is not considered
a practical alternative.
Transshipment of the Fuel Offsite to Another FPC Site
CR-3 is the only nuclear unit of FPC. Therefore, transshipment of
spent fuel to another facility with FPC is not an available option.
Decommissioning
Power generation from CR-3 is essential to meet the current growth
rate for energy demand in the State of Florida. Additional replacement
capacity would be required if CR-3 were to be retired early. Permanent
shutdown of CR-3 would result in loss of valuable power resources. The
environmental impact would be similar to that for operation at a
reduced power level.
Alternatives Creating Additional Storage Capacity
Dry cask storage is a method of transferring spent fuel, after
storage in the pool for several years, to high-capacity casks with
passive heat dissipation features. Storage of fuel in a private
Independent Spent Fuel Storage Installation (ISFSI) located away from
the CR-3 site is not available, since such a facility has not been
constructed by FPC or licensed by the NRC. An on-site
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ISFSI is a long-term solution for CR-3, but cost and schedule
considerations do not allow this alternative to meet current needs at
CR-3 for near term spent fuel storage needs.
The alternative technology of constructing an ISFSI that could
create additional storage capacity involves additional fuel handling
with an attendant opportunity for a fuel handling accident, involves
higher cumulative dose to workers affecting the fuel transfers, and
would not result in a significant improvement in environmental impacts
compared to the proposed reracking modifications.
The No-Action Alternative
The NRC staff also considered denial of the proposed action (i.e.,
the ``no-action'' alternative). Denial of the application would result
in no significant change in current environmental impacts. The
environmental impacts of the proposed action and the alternative
actions are similar.
Alternative Use of Resources
This action does not involve the use of any resources not
previously considered in the FES for CR-3.
Agencies and Persons Contacted
In accordance with its stated policy, on August 7, 2000, the NRC
staff consulted with William Passetti, Chief, Department of Health,
Bureau of Radiation Control, for the State Florida, regarding the
environmental impact of the proposed action. The state official had no
comments.
Finding of No Significant Impact
On the basis of the environmental assessment, the NRC concludes
that the proposed action will not have a significant effect on the
quality of the human environment. Accordingly, the NRC 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 September 16, 1999, as supplemented by letters
dated May 3 and June 29, 2000, which are available for public
inspection at the Commission's Public Document Room, The Gelman
Building, 2120 L Street, NW., Washington, DC. Publicly available
records will be accessible electronically from the ADAMS Public Library
Component on the NRC Web site, http://www.nrc.gov.
Dated at Rockville, Maryland, this 5th day of September 2000.
Richard P. Correia,
Chief, Section 2, Project Directorate II, Division of Licensing Project
Management, Office of Nuclear Reactor Regulation.
[FR Doc. 00-23357 Filed 9-11-00; 8:45 am]
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