[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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