[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

[[Page 55437]]

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