[Federal Register Volume 73, Number 37 (Monday, February 25, 2008)]
[Notices]
[Pages 10069-10071]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E8-3486]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-395]
South Carolina Electric & Gas Company,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 exemption from Title 10 of the Code of Federal
Regulations, Part 50, (10 CFR), Section 50.46, ``Acceptance criteria
for emergency core cooling systems for light-water nuclear power
reactors,'' (10 CFR 50.46) and 10 CFR Part 50, Appendix K, ``ECCS
Evaluation Models,'' (Appendix K) for the Renewed 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.
Therefore, as specified in 10 CFR 51.21, the NRC has performed an
environmental assessment as described in this notice and has made a
finding of no significant impact.
The action proposed by the licensee also included a request for an
exemption from 10 CFR 50.44, ``Combustible gas control for nuclear
power reactors,'' (10 CFR 50.44). The proposed exemption from 10 CFR
50.44 is not being considered further by the NRC staff because
revisions to 10 CFR 50.44 (68 FR 54123, dated September 16, 2003), such
that it does not refer to specific types of zirconium cladding, remove
the need for such an exemption.
Environmental Assessment
Identification of the Proposed Action
The proposed action would allow a third cycle of irradiation (i.e.,
burnup) for one lead test assembly (LTA) containing fuel rods with
advanced cladding alloys. This third cycle of irradiation is expected
to begin in the Cycle 18 core for VCSNS in the spring of 2008. An
exemption previously issued by the NRC on January 14, 2005, authorized
the use of four LTAs up to a lead rod average burnup limit of 62,000
megawatt days per metric ton uranium (MWd/MTU). The cladding in two of
those four LTAs is entirely Optimized ZIRLOTM cladding. Each
of the other two LTAs uses sixteen fuel rods with AXIOMTM
cladding with the remainder of the rods using Optimized
ZIRLOTM cladding. Based upon the results of examinations of
these four LTAs during the VCSNS Cycle 17/18 refueling outage, the
licensee may select either one of the Optimized ZIRLOTM LTAs
or one of the LTAs containing both Optimized ZIRLOTM plus
AXIOMTM cladding for the third cycle of irradiation. The
third cycle of irradiation is expected to take the LTA from a burnup of
about 55,000 up to 75,000 MWd/MTU. The burnup limits are not part of
the technical specifications (TS), but are design bases limits, and
limit the current fuel rod-average burnup to less than or equal to
62,000 MWd/MTU. The proposed action is in accordance with the
licensee's application dated May 31, 2007, as supplemented by letter
dated October 11, 2007. Also, information in the licensee's letters
dated September 3 and November 11, 2004, that supported the exemption
previously issued on January 14, 2005, has been considered in this
action.
The Need for the Proposed Action
As the licensee states in its letter dated September 3, 2004, ``As
the nuclear industry pursues longer operating cycles with increased
fuel discharge burnups and more aggressive fuel management, corrosion
performance requirements for nuclear fuel cladding become more
demanding. In addition, fuel rod internal pressures (resulting from
increased fuel duty, use of integral fuel burnable absorbers (IFBAs)
and corrosion/temperature feedback effects) have become more limiting
with respect to fuel rod design criteria. Available industry data [* *
*] indicate the corrosion resistance improves for cladding with a lower
tin content,'' and ``In addition, developmental testing has shown that
small additions of some alloying elements will further improve the
corrosion resistance, microstructure and mechanical properties of the
cladding,'' and ``To meet these needs, Westinghouse Electric Company
has developed a lead test assembly program in cooperation with the V.C.
Summer Nuclear Station. One element of the program is use of Optimized
ZIRLOTM cladding [* * *]'' and another element of the
program is the use of LTAs with AXIOMTM cladding.
As the licensee states in its application, 10 CFR 50.46
specifically refers to fuel with Zircaloy or ZIRLOTM
cladding and does not include Optimized ZIRLOTM or
AXIOMTM cladding. Appendix K, paragraph I.A.5, references an
analysis that utilizes the Baker-Just equation which assumes use of a
zirconium alloy different than the Optimized ZIRLOTM or
AXIOMTM cladding used in the LTAs. Therefore, the exemption
is needed because the NRC regulations identified above specifically
refer to light-water reactors
[[Page 10070]]
containing fuel consisting of uranium oxide pellets enclosed in
zircaloy or ZIRLOTM cladding and the newer zirconium-based
alloys of Optimized ZIRLOTM and AXIOMTM are not
specifically of the same composition as zircaloy or ZIRLOTM.
Therefore, the licensee needs an exemption to insert one of the four
above-mentioned LTAs into the VCSNS reactor core for further
irradiation.
Environmental Impacts of the Proposed Action
The NRC has completed its evaluation of the proposed action and
concludes that there are no significant environmental impacts
associated with the use of one fuel assembly using either all Optimized
ZIRLOTM cladding or a combination of Optimized
ZIRLOTM and AXIOMTM cladding for a third cycle of
irradiation up to a burnup of 75,000 MWd/MTU. The following is a
summary of the staff's evaluation:
In this environmental assessment, the NRC staff is also relying on
the results of a study conducted for it by the Pacific Northwest
National Laboratory (PNNL) entitled, ``Environmental Effects of
Extending Fuel Burnup Above 60 GWd/MTU [gigawatt days per metric ton
uranium],'' (NUREG/CR-6703, PNNL-13257, January 2001). Although the
study evaluated the environmental impacts of high burnup fuel up to
75,000 MWd/MTU, certain aspects of the review were limited to
evaluating the impacts of extended burnup up to 62,000 MWd/MTU because
of the need for additional data about the effect of extended burn-up on
gap-release fractions. During the study, all aspects of the fuel-cycle
were considered, from mining, milling, conversion, enrichment and
fabrication through normal reactor operation, transportation, waste
management, and storage of spent fuel.
The staff has concluded that such changes would not adversely
affect plant safety, and would have no adverse effect on the
probability of any accident. For accidents that involve damage or
melting of the fuel in the reactor core, fuel rod integrity has been
shown to be unaffected by the extended burnup under consideration;
therefore, the probability of an accident will not be affected. For
accidents in which the core remains intact, the increased burnup may
slightly change the mix of fission products that could be released in
the event of a serious accident, however the staff concludes that the
limited number of high burnup fuel rods in one LTA will not result in a
significant change during core-wide events.
Accidents that involve the damage or melting of the fuel in the
reactor core and spent-fuel handling accidents were also evaluated in
NUREG/CR-6703. The accidents considered were a loss-of-coolant accident
(LOCA), a steam generator tube rupture, and a fuel-handling accident.
For LOCAs, the amount of radionuclides that would be released from
the core (1) is proportional to the amount of radionuclides in the core
and (2) is not significantly affected by the gap-release fraction. The
gap-release fraction is a small contributor to the amount of
radionuclides available for release when the fuel is severely damaged.
Any increase in the amount of some longer-lived radionuclides available
for release from the single LTA (1) will be small and (2) will not
result in a significant increase in the overall core inventory of
radionuclides. Therefore, there would be no significant increase in the
previously calculated dose from a LOCA and the dose would remain below
regulatory limits.
The pressurized-water reactor steam generator tube rupture accident
involves direct release of radioactive material from contaminated
reactor coolant to the environment. No change is being requested by the
licensee in the VCSNS TS pertaining to allowed cooling-water activity
concentrations. The maximum coolant activity is regulated through TS
that are independent of fuel burnup. Therefore, the gap-release
fraction does not significantly affect the amount of radionuclides
available for release during a steam generator tube rupture. Therefore,
there would be no significant increase in the previously calculated
dose from a steam generator tube rupture and the calculated dose would
remain below regulatory limits.
The scenario postulated to evaluate potential fuel-handling
accidents involves a direct release of gap activity to the environment.
The assumptions regarding gap activity are based on guidance in
Regulatory Guide 1.25, ``Assumptions Used for Evaluating the Potential
Radiological Consequences of Fuel Handling Accidents in the Fuel
Handling and Storage Facility for Boiling and Pressurized Water
Reactors (Safety Guide 25)'' and NUREG/CR-5009, ``Assessment of the Use
of Extended Burnup Fuel in Light Water Power Reactors,'' February 1988;
the gap activity consists primarily of noble gases and iodine. The
isotopes that contribute significant fractions of the whole body and
thyroid doses are 87Kr and 131I, respectively.
The inventory of iodine, the primary dose contributor, decreases with
increasing burnup. In addition, the single LTA will only contribute a
small variation in the isotopic population of the entire VCSNS core
(157 assemblies).
The licensee assessed, in its letter dated October 11, 2007, the
conservatisms associated with the spent fuel pool decontamination
factor, the assembly relative power, the thyroid dose conversion
factors, fuel offloading time, the reactor building purge isolation and
the likely mechanical damage to a fuel assembly from the fuel handling
accident. In summarizing these factors the licensee estimates that the
calculated doses for the fuel handling accident would be reduced by
approximately 77 percent. Based on the considerations discussed above,
the staff concludes (1) that the increase in the previously calculated
dose resulting from a fuel-handling accident involving the one LTA
would not be significant and (2) that the dose would remain below
regulatory limits.
Regulatory limits on radiological effluent releases are independent
of burnup. The requirements of 10 CFR 50.36a and Appendix I to 10 CFR
part 50 ensure that any release of gaseous, liquid, or solid
radiological effluents to unrestricted areas are kept ``as low as
reasonably achievable.'' Therefore, the staff concluded that during
routine operations, there will be no significant increase in the amount
of gaseous radiological effluents released into the environment as a
result of the proposed action, nor will there be a significant increase
in the amount of liquid radiological effluents or solid radiological
effluents released into the environment.
No significant increase in the allowable individual or cumulative
occupational radiation exposure will occur. The impacts to workers is
expected to be reduced with higher irradiation due to the need for less
frequent outages for fuel changes and less frequent fuel shipments to
and from reactor sites.
The use of extended irradiation will not change the potential
environmental impacts of incident-free transportation of spent nuclear
fuel or the accident risks associated with spent fuel transportation if
the fuel is cooled for 5 years after discharge from the reactor. The
NUREG/CR-6703 report, concluded that doses associated with incident-
free transportation of spent fuel with burnup to 75 GWd/MTU are bounded
by the doses given in 10 CFR 51.52, Table S-4, for all regions of the
country if dose rates from the shipping casks are maintained within
regulatory limits. Increased fuel burnup will decrease the annual
discharge of fuel to the spent fuel pool, which will postpone the need
to remove spent fuel from the pool.
With regard to potential non-radiological environmental impacts of
[[Page 10071]]
reactor operation with extended irradiation, the proposed changes
involve systems located within the restricted area as defined in 10 CFR
part 20. Therefore, the proposed action does not result in any
significant changes to land use or water use, or result in any
significant changes to the quality or quantity of effluents. The
proposed action does not affect nonradiological plant effluents, and no
changes to the National Pollution Discharge Elimination System permit
are needed. No effects on the aquatic or terrestrial habitat in the
vicinity or the plant, or to endangered or threatened species, or to
the habitats of endangered or threatened species are expected. The
proposed action does not have a potential to affect any historical or
archaeological sites.
The proposed action will not change the method of generating
electricity or the method of handling any influents from the
environment or non-radiological effluents to the environment.
Therefore, no changes or different types of non-radiological
environmental impacts are expected as a result of the amendments.
Accordingly, the NRC concludes that there are no significant
environmental impacts associated with the proposed action.
For more detailed information regarding the environmental impacts
of extended fuel burnup, please refer to the study conducted by PNNL
for the NRC, which is entitled, ``Environmental Effects of Extending
Fuel Burnup Above 60 GWd/MTU,'' (NUREG/CR-6703, PNL-13257, January
2001).
The details of the staff's safety evaluation will be provided in
the exemption that will be issued as part of the letter to the licensee
approving the exemption to the regulation.
Environmental Impacts of the Alternatives to the Proposed Action
As an alternative to the proposed action, the staff considered
denial of the proposed action (i.e., the ``no-action'' alternative).
Denial of the amendment request would result in no change in current
environmental impacts. The environmental impacts of the proposed
amendment and this alternative are similar. However, it would deny to
the licensee and the NRC operational data on Optimized
ZIRLOTM and AXIOMTM LTAs and the performance of
fuel at extended burnup conditions.
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
the Virgil C. Summer Nuclear Station, NUREG-0719, dated May 1981, or in
NUREG-1437, Supplement 15, ``Generic Environmental Impact Statement for
License Renewal of Nuclear Plants, Supplement 15, Regarding Virgil C.
Summer Nuclear Station.''
Agencies and Persons Consulted
In accordance with its stated policy, on December 31, 2007, the
staff consulted with the South Carolina State official, R. Mike Gandy
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 May 31, 2007 (Agencywide Documents Access and
Management System (ADAMS) Accession No. ML071550105), as supplemented
on October 11, 2007 (ADAMS Accession No. ML072890083). Documents may be
examined, and/or copied for a fee, at the NRC's Public Document Room
(PDR), located at One White Flint North, 1555 Rockville Pike,
Rockville, Maryland 20852. Publicly available records will be
accessible electronically from the 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 send an e-mail to [email protected].
Dated at Rockville, Maryland, this 12th day of February, 2008.
For the Nuclear Regulatory Commission.
Robert Martin,
Project Manager, Plant Licensing Branch II-1, Division of Operating
Reactor Licensing, Office of Nuclear Reactor Regulation.
[FR Doc. E8-3486 Filed 2-22-08; 8:45 am]
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