[Federal Register Volume 67, Number 168 (Thursday, August 29, 2002)]
[Notices]
[Pages 55436-55439]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-22108]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-395]
South Carolina Electric & Gas Co.; Virgil C. Summer Nuclear
Station; Environmental Assessment and Finding of No Significant Impact
The U.S. Nuclear Regulatory Commission (NRC) is considering
issuance of an amendment to Title 10 of the Code of Federal Regulations
(10 CFR) part 50, Sec. 50.90 for Facility Operating License No. NPF-12,
issued to South Carolina Electric & Gas Company (SCE&G, the licensee),
for operation of the Virgil C. Summer Nuclear Station (VCSNS), located
in Fairfield County, South Carolina. As required by 10 CFR 51.21, the
NRC is issuing this environmental assessment and finding of no
significant impact.
Environmental Assessment
Identification of the Proposed Action
The proposed action would increase the spent fuel pool (SFP)
storage capacity by replacing all 11 existing rack modules with 12 new
storage racks. The rerack will increase the storage capacity from 1,276
storage cells to 1,712 storage cells. The new racks will have Boral
neutron-absorbing material instead of the degrading Boraflex used in
the existing racks.
The proposed action is in accordance with the licensee's
application dated July 24, 2001, as supplemented by letters dated April
4, 2002, May 7, 2002, June 17, 2002, July 2, 2002, July 15, 2002, and
July 25, 2002.
The Need for the Proposed Action
SCE&G currently expects VCSNS to lose the capacity for full-core
offload during refueling operations in 2008 (after Cycle 17). SCE&G has
evaluated spent fuel storage options that have been licensed by the NRC
and are currently feasible for use at the VCSNS site. The evaluation
concluded that reracking the SFP is currently the most cost-effective
alternative. Reracking would increase storage capacity and maintain the
plant's capability to accommodate a full-core discharge until the end
of Cycle 24 in 2018.
Environmental Impacts of the Proposed Action
Solid Radioactive Waste
Spent resins are generated by the processing of SFP water through
the SFP purification system. The licensee predicts that the
installation of the new racks will generate slightly more resin from
the new, increased capacity rack installation; therefore, the licensee
may more frequently change-out the SFP purification system during the
reracking operation. In order to keep the SFP water reasonably clear
and clean and thereby minimize the generation of spent resins, the
licensee will vacuum the floor of the SFP as necessary to remove any
radioactive crud, sediment, and other debris before the new fuel rack
modules are installed. The filters from this underwater vacuum will be
a minor source of solid radioactive waste. However, the licensee does
not expect that the increase in storage capacity of the SFP will result
in a significant change in the long-term generation of solid
radioactive waste at VCSNS.
The disposal of the used spent fuel racks will result in a one-time
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incremental increase in solid waste. Because ongoing volume reduction
efforts have effectively minimized the amount of waste generated, this
incremental 1-year increase is bounded by the plant's original
licensing basis described in the Final Environmental Statement for the
VCNS (NUREG-0719) dated May 1981, and therefore is acceptable.
Gaseous Radioactive Waste
The storage of additional spent fuel assemblies in the SFP is not
expected to affect the releases of radioactive gases from the SFP.
Gaseous fission products such as krypton-85 and iodine-131 are produced
by the fuel in the core during reactor operation. Small amounts of
these fission gases are released to the reactor coolant from the small
number of fuel assemblies that develop leaks during reactor operation.
During refueling operations, some of these fission products enter the
SFP and are subsequently released into the air. Since the frequency of
refuelings, and therefore the number of freshly off-loaded spent fuel
assemblies stored in the SFP at any one time, will not increase, there
will be no increase in the amounts of gaseous fission products released
to the atmosphere as a result of the increased SFP fuel storage
capacity.
The increased heat load on the SFP from the storage of additional
spent fuel assemblies could potentially increase the SFP evaporation
rate. However, based on previous reracks at other facilities, this
increased evaporation rate is not expected to significantly increase
the amount of gaseous tritium released from the pool. Thus, the
licensee does not expect the concentrations of airborne radioactivity
in the vicinity of the SFP to significantly increase due to the
expanded SFP storage capacity. This is consistent with the operating
experience to date with previous SFP expansions. Gaseous effluents from
the spent fuel storage area are combined with other station exhausts
and monitored before release. Past SFP area contributions to the
overall site gaseous releases have been insignificant and should remain
negligible with the increased capacity. The impact of any increases in
site gaseous releases should be negligible, and the resultant doses to
the public will remain very small fractions of the 10 CFR part 20 and
10 CFR part 50, appendix I, dose limits.
Liquid Radioactive Waste
The release of radioactive liquids will not be affected directly as
a result of the SFP expansion. The SFP ion exchanger resins remove
soluble radioactive materials from the SFP water. When the resins are
changed out, the small amount of resin sluice water is processed by the
radioactive waste system before release to the environment. As stated
above, the frequency of resin change out may increase slightly during
the installation of the new racks. However, the increase in the amount
of liquid effluents released to the environment as a result of the
proposed SFP expansion is expected to be negligible.
Occupational Radiation Exposure
The NRC staff has reviewed the licensee's plan for the modification
of the VCSNS spent fuel racks with respect to occupational radiation
exposure. As stated above, the licensee plans to remove the 11 existing
fuel racks and install 12 new racks in the SFP. Based on the lessons
learned from a number of facilities that have performed similar
operations in the past and their experience with reracks, the licensee
estimates that the collective occupational worker dose for the proposed
fuel rack project will be between 6 and 12 person-rem.
All of the operations involved in the removal of existing racks and
the installation of the new fuel racks will be governed by procedures.
These procedures are based on the principle of keeping doses as low as
reasonably achievable (ALARA), consistent with the requirements of 10
CFR part 20. The radiation protection department will prepare a
radiation work permit (RWP) for the various in-pool and out-of-pool
jobs. The RWP and supporting job procedures will establish requirements
for timely external radiation and airborne surveys, personal protective
clothing and equipment, individual monitoring devices, and other access
and work controls consistent with good radiation protection practices
and 10 CFR part 20 requirements. Continuous health physics technician
(HPT) coverage will be provided and maintained when a diver is in the
pool, and when any potentially contaminated object is being removed
from the pool. Each member of the project team will receive radiation
protection training on the reracking operations consistent with the
requirements of 10 CFR part 19. Project-specific training will include
hot particle hazards and the potential for extremity doses from working
in the fuel pool or with the old racks (e.g., decontaminating and
packaging them for shipment off-site). Prior to the start of the job,
lessons learned from previous pool rerackings will be discussed as part
of the ALARA briefing. Daily pre-job briefings, which will include
information on pertinent ALARA issues, will be used to inform workers
and HPTs of job scope and techniques. All divers will be fully trained
and qualified for nuclear diving.
For out-of-pool work activities, all workers will be provided with
thermoluminescence dosimeters (TLDs) and electronic alarm dosimeters.
Additional personal monitoring devices (e.g., extremity badges) will be
used, as appropriate. Periodic radiation surveys will be conducted for
direct radiation levels and loose surface contamination levels, as
appropriate and in accordance with the governing RWP. Historical
experience during similar reracking shows that radioactive airborne
material levels in the above-pool work area should be negligible during
the rerack job. However, air sampling will be performed, and continuous
air monitors will be used, when a job evolution has the potential for
generating significant airborne radioactivity. Personal respiratory
equipment will be available, if needed. In order to minimize
contamination and airborne problems, all equipment removed from the
pool will be surveyed before removal, surveyed as it breaks the water
surface, rinsed off and wiped down, and resurveyed by or under the
direction of a qualified HPT.
The VCSNS SFP rerack project will use qualified underwater divers
for both rack removal and installation. No divers will be allowed in
the SFP during any movement of spent fuel to ensure that these divers
are not exposed to high and very high radiation sources (e.g., spent
fuel). All diving operations will be governed by special procedures.
These procedures will require extensive surveys of the dive area before
dives and divers will be trained to use calibrated underwater radiation
survey instruments for confirmatory surveys of their work area. The
location of significant radiation sources will be made known to the
divers, and the divers' range of motion in the SFP will be restricted
by a tether, which will help ensure that a diver does not get too close
to high and very high radiation sources. Additionally, underwater
barriers will be used to physically define the safe dive area. No
deviations from the planned, prescribed dive will be allowed.
Continuous audio and video monitoring and communication will be in
place to allow for constant poolside surveillance of all diver
activities. If any of these monitoring capabilities are lost, the dive
will be terminated. Each diver will be provided with multiple TLDs and
electronic dosimeters for whole body and extremity monitoring, with
continuous remote dose rate readouts for poolside observation,
monitoring,
[[Page 55438]]
and control, because of the steep dose gradients in the water
shielding. The VCSNS diving control and survey procedures described
above meet the intent of Regulatory Guide 8.38, ``Control of Access to
High and Very High Radiation Areas in Nuclear Power Plants,'' Appendix
A, ``Procedures for Diving Operations in High and Very High Radiation
Areas.'' This appendix was developed from the lessons learned from
previous diver overexposures and mishaps, and summarizes good operating
practices for divers acceptable to the NRC staff.
An underwater vacuum system will be used to supplement the
installed SFP filtration system so that the levels of radiation and
contamination, including hot particles and debris, can be reduced
before diving operations. The SFP floor dive area will be vacuum-
cleaned with long-handled tools from above the pool. Final radiation
surveys and visual inspection by underwater camera will be performed
before any diving activities. These actions to identify and control hot
particles and debris should effectively minimize the potential for
unplanned diver exposures from these sources.
Before the old fuel racks are removed from the pool, they will be
cleaned underwater using high-pressure washing. After cleaning, while
the racks are still over the pool, radiation surveys will be performed
to determine if further decontamination is needed before the racks are
prepared for shipment off-site. The racks will be bagged remotely to
minimize potential worker contamination and maintain doses ALARA. Once
properly packaged in approved shipping containers, the racks will be
shipped in accordance with Department of Transportation and NRC
regulations. The licensee will use the existing SFP filtration system
during fuel rack installation to maintain water clarity in the SFP.
These engineering controls and handling procedures will help minimize
the spread of contamination (e.g., hot particles), while keeping worker
doses ALARA.
The storage of additional spent fuel assemblies in the SFP, and the
reduction in minimum cooling time from 100 hours down to 72 hours
before fuel movement, will result in negligible increases in the
external dose rates on the refueling floor and in accessible areas
adjacent to the SFP. Existing normally accessible areas around the fuel
storage pool are designated Radiation Zone II. That designation will be
maintained with the external dose rates remaining less than 2.5 mrem/
hr. The maximum dose rates outside the concrete walls of the SFP will
remain less than 0.01 mrem/hr. The area most impacted by the pool
rerack is the fuel transfer canal (FTC), assuming it to be drained and
empty. Assuming an empty FTC, to keep radiation levels below 2.5 mrem/
hr, procedures will require that no fuel except old fuel be stored near
the gate slot to the FTC. Normally, the FTC will be filled with water.
On the basis of our review of the VCSNS proposal, the NRC staff
concludes that the SFP rerack can be performed in a manner that will
ensure that doses to the workers will be maintained ALARA. The NRC
staff finds the projected dose for the project of about 6 to 12 person-
rem to be appropriate and in the range of doses for similar SFP
modifications at other plants, and therefore acceptable.
Fuel Handling Accident (FHA) Radiological Consequences
The design-basis FHA analysis postulates that a spent fuel assembly
is dropped during refueling, damaging all of the rods in the assembly
plus 50 additional rods in an adjacent assembly (a total of 314 rods).
The design of the fuel handling equipment makes it very likely that a
dropped assembly would result in the release of fission products. The
accident analysis assesses whether design features for mitigating
environmental releases meet certain design criteria. At VCSNS, this
accident could happen inside the containment (CNMT) or in the fuel
handling building (FHB), and SCE&G has evaluated both cases.
The SCE&G analyses assume that core inventory is based on 5-percent
by weight initial enrichment fuel and extended operation at 2958 MWt
power. The core inventory was determined using the NRC-sponsored SCALE
computer code suite. SCE&G considered five fuel burnup exposures
ranging from 35,000 MWt/MTU to 70,000 MWt/MTU. (This assessment does
not address operation above a burnup of 62,000 MWt/MTU.) Since
individual radionuclides reach peak equilibrium values at different
rates, the highest specific inventory of each contributing radionuclide
in any of the burnup ranges was used in the analyses. A decay period of
72 hours between reactor shutdown and fuel movement was assumed. Since
the power level and, hence, the inventory in each assembly varies
across the core, a radial peaking factor of 1.7 is applied to the
average core inventory. SCE&G assumed that 12 percent of the I-131
inventory of the core was in the fuel rod gap, along with 30 percent of
the Kr-85, and 10 percent of all other iodines and noble gases. The
radioiodine in the gap was assumed to be 99.75 percent elemental and
0.25 percent organic forms.
SCE&G assumes that all of the gap inventory in the 314 damaged fuel
rods is instantaneously released through the water in the reactor
cavity or SFP into the CNMT or FHB, respectively. SCE&G assumes that
100 percent of the activity release to the CNMT or FHB is released to
the environment in 2 hours. Credit was taken for the FHB purge exhaust
charcoal filters, but no credit was taken for the reactor building
purge exhaust charcoal filters.
Details on the assumptions found acceptable to the NRC staff are
presented in the attached Table. The offsite doses estimated by the
licensee for the postulated FHAs were found to be acceptable.
The NRC has completed its evaluation of the proposed action and
concludes that the proposed action will not significantly increase the
probability or consequences of accidents, no changes are being made in
the types of effluents that may be released off site, and there is no
significant increase in occupational or public radiation exposure. The
incremental 1-year increase in waste is bounded by the plant's original
licensing basis and is therefore acceptable. Therefore, there are no
significant radiological environmental impacts associated with the
proposed action.
With regard to potential nonradiological impacts, the proposed
action does not have a potential to affect any historic sites. It does
not affect nonradiological plant effluents and has no other
environmental impact. Therefore, there are no significant
nonradiological environmental impacts associated with the proposed
action.
Accordingly, the NRC concludes that there are no significant
environmental impacts associated with the proposed action.
Environmental Impacts of the Alternatives to the Proposed Action
According to Holtec Report HI-20112624, ``Fuel Storage Expansion at
Virgil C. Summer for South Carolina Electric & Gas,'' the following
alternative actions were considered:
Rod Consolidation
Rod consolidation has been shown to be a potentially feasible
technology. Rod consolidation involves disassembly of one [fuel
assembly] and the disposal of the fuel assembly skeleton outside of
the pool (this is considered a 2:1 compaction ratio). The rods are
stored in a stainless steel can that has the outer dimensions of a
fuel assembly. The can is stored in the spent fuel racks. The top of
the can has an end fixture that matches up
[[Page 55439]]
with the spent fuel handling tool. This permits moving the cans in
an easy fashion.
Rod consolidation pilot project campaigns in the past have
consisted of underwater tooling that is manipulated by an overhead
crane and operated by a maintenance worker. This is a very slow and
repetitive process.
The industry experience with rod consolidation has been mixed
thus far. The principal advantages of this technology are: The
ability to modularize, compatibility with the U.S. Department of
Energy (DOE) waste management system, moderate cost, no need of
additional land and no additional required surveillance. The
disadvantages are: potential gap activity release due to rod
breakage; potential for increased fuel cladding corrosion due to
some of the protective oxide layer being scraped off; potential
interference of the (prolonged) consolidation activity, which might
interfere with ongoing plant operation; and lack of sufficient
industry experience. The drawbacks associated with consolidation are
expected to diminish in time. However, it is the SCE&G's view that
rod consolidation technology has not matured sufficiently to make
this a viable option for the present VCSNS spent fuel pool
limitations.
On-Site Dry Cask Storage
Dry cask storage is a method of storing spent nuclear fuel in a
high capacity container. The cask provides radiation shielding and
passive heat dissipation. Typical capacities for pressurized-water
reactor fuel range from 21 to 37 assemblies that have been removed
from the reactor for at least 5 years. The casks, once loaded, are
then stored outdoors on a seismically qualified concrete pad.
The casks, as presently licensed, are limited to 20-year storage
service life. Once the 20 years has expired, the cask manufacturer
or the utility must recertify the cask or the utility must remove
the spent fuel from the container. In the interim, DOE has embraced
the concept of multi-purpose canisters obsolescing all existing
licensed cask designs. Work is also continuing by several companies,
including Holtec International, to provide an [a] multi-purpose
canister system that will be capable of long storage, transport, and
final disposal in a repository. Holtec International's HI-STAR
System can store up to 24 pressurized-water reactor assemblies. It
is noted that a cask system makes substantial demands on the
resources of a plant. For example, the plant must provide for a
decontamination facility where the outgoing cask can be
decontaminated for release.
There are several plant modifications required to support cask
use. Tap-ins must be made to the gaseous waste system, and chilled
water to support vacuum drying of the spent fuel and piping must be
installed to return cask water back to the Spent Fuel Pool/Cask
Loading Pit. A seismic concrete pad must be made to store the loaded
casks. This pad must have a security fence, surveillance protection,
a diesel generator for emergency power, and video surveillance for
the duration of fuel storage, which may extend beyond the life of
the adjacent plant. Finally, the cask park must have facilities to
vacuum dry the cask, backfill it with helium, make leak checks,
remachine the gasket surfaces if leaks persist, and assemble the
cask on-site.
To summarize, based on the required short time schedule, the
status of the dry spent fuel storage industry, and the storage
expansion costs, the most acceptable alternative for increasing fuel
storage capacity at VCSNS is expansion of the wet storage capacity.
No-Action Alternative
As an alternative to the proposed action, the staff considered
denial of the proposed action (i.e., the ``no-action'' alternative).
Denial of the application would result in no change in current
environmental impacts. The environmental impacts of the proposed action
and the alternative actions are similar.
The alternative technologies that could create additional storage
capacity involve additional fuel handling with increased opportunity
for fuel handling accidents, involve higher commutative doses to
workers affecting the fuel transfers and would not result in a
significant improvement in environmental impacts compared to the
proposed reracking modifications.
Alternative Use of Resources
The action does not involve the use of any different resources than
those previously considered in the Final Environmental Statement for
VCSNS (NUREG-0719) dated May 1981.
Agencies and Persons Consulted
On July 23, 2002, the staff consulted with the South Carolina State
official, Mr. Henry Porter of the South Carolina Department of Health
and Environmental Control, 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 July 24, 2001, and supplemental letters dated
April 4, 2002, May 7, 2002, June 17, 2002, July 2, 2002, July 15, 2002,
and July 25, 2002. Documents may be examined, and/or copied for a fee,
at the NRC's Public Document Room (PDR), located at One White Flint
North, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible electronically from the
Agencywide Documents Access and Management System (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to
ADAMS or who encounter problems in accessing the documents located in
ADAMS should contact the NRC PDR Reference staff by telephone at 1-800-
397-4209 or 301-415-4737, or by e-mail to [email protected].
Dated at Rockville, Maryland, this 26th day of August, 2002.
For the Nuclear Regulatory Commission.
John A. Nakoski,
Chief, Section 1, Project Directorate II, Division of Licensing Project
Management, Office of Nuclear Reactor Regulation.
[FR Doc. 02-22108 Filed 8-28-02; 8:45 am]
BILLING CODE 7590-01-P