[Federal Register Volume 69, Number 132 (Monday, July 12, 2004)]
[Notices]
[Pages 41852-41855]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-15696]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-413 and 50-414]


Duke Energy Corporation, et al., Catawba Nuclear Station, Units 1 
and 2; Notice of Opportunity To Comment and Proposed No Significant 
Hazards Consideration Determination

    The U.S. Nuclear Regulatory Commission (the Commission) is 
reviewing an application for amendment to Facility Operating License 
Nos. NPF-35 and NPF-52, issued to Duke Power Company, et al. (the 
licensee), for operation of the Catawba Nuclear Station (Catawba), 
Units 1 and 2, located in York County, South Carolina. A Notice of 
Consideration of Issuance of Amendment to Facility Operating License 
and Opportunity for a Hearing was published in the Federal Register on 
July 25, 2003.
    The proposed amendments, requested by the licensee in a letter 
dated February 27, 2003, as supplemented by letters dated September 15, 
September 23, October 1 (two letters), October 3 (two letters), 
November 3 and 4, December 10, 2003, February 2, 2004, (two letters), 
March 1 (two letters), March 9 (two letters), and March 16, (two 
letters), March 26, March 31, April 13, April 16, May 13 and June 17, 
2004, would revise the Technical Specifications to allow the use of 
four mixed oxide (MOX) fuel lead test assemblies (LTAs). The term 
``MOX'' arises from the following: the low enriched uranium (LEU) fuel 
used in U.S. reactors heretofore consists mostly of uranium oxides 
wherein the concentration of U-235 is increased during manufacture, 
such that U-235 constitutes up to four to five percent of the uranium 
by weight. In fresh unirradiated LEU fuel, U-235 is the fissionable 
component and it has no significant plutonium concentration. During 
irradiation, however, U-238 absorbs neutrons produced by the fission of 
U-235 and transmutes to the various isotopes of plutonium. Some of 
these plutonium isotopes are fissionable

[[Page 41853]]

and add to the power output of the LEU fuel such that with the 
irradiation of LEU fuel to medium to high burnup levels, a significant 
fraction of that fuel's power is produced by the fissioning of 
plutonium. As a part of a joint United States-Russian surplus weapons-
grade plutonium disposition program supported by the Department of 
Energy (DOE) to reduce the threat of nuclear weapons proliferation 
worldwide by conducting disposition of surplus plutonium in the United 
States, the licensee proposes that plutonium oxide powder supplied by 
DOE will be processed, blended with depleted uranium dioxide powder, 
and fabricated into MOX fuel LTAs that will then be used at Catawba. 
The blending of the uranium oxide and plutonium oxide materials is the 
basis for the term ``mixed oxide'' or ``MOX'' fuel.
    Before issuance of the proposed license amendments, the Commission 
will have made findings required by the Atomic Energy Act of 1954, as 
amended (the Act) and the Commission's regulations.
    The Commission has made a proposed determination that the amendment 
request involves no significant hazards consideration. Under the 
Commission's regulations in Title 10 of the Code of Federal Regulations 
(10 CFR) Section 50.92, this means that operation of the facility in 
accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated; or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. As required 
by 10 CFR 50.91(a), the licensee has provided its analysis of the issue 
of no significant hazards consideration, parts of which are presented 
below.

I. Probability and Consequences Evaluation

    The proposed license amendment to allow the use of MOX fuel lead 
assemblies does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The ``accidents'' previously evaluated are described in the 
[Updated Final Safety Analysis Report] UFSAR and fall into one of the 
following four categories:
     Normal Operation and Operational Transients.
     Faults of Moderate Frequency.
     Infrequent Faults.
     Limiting Faults.
    Inspection of the UFSAR descriptions reveals that the presence of 
MOX fuel lead assemblies could potentially impact the probability of 
occurrence for only two ``accidents;'' Radioactivity in Reactor Coolant 
Due to Cladding Defects and Fuel Handling Accidents. An evaluation of 
each of these events follows.

Radioactivity in Reactor Coolant Due to Cladding Defects Probability

    Cladding defects are imperfections in the cladding material of a 
fuel assembly that allow fission products from the active fuel material 
to migrate to the reactor coolant. They can be caused by manufacturing 
defects that go undetected until the stresses of pressure, temperature, 
and/or irradiation eventually result in fuel cladding failure. This 
type of cladding failure occurs very infrequently in low-enriched 
uranium (LEU) fuel. The Mark BW design, which is the basis for the Mark 
BW/MOX1 design to be used in the MOX fuel lead assemblies, has 
experienced a failure rate of less than one per 100,000 rods, from all 
manufacturing related causes, since its inception in 1987. There is no 
reason to expect that the probability of this type of failure in a MOX 
fuel assembly will be any different than for a LEU fuel assembly 
because the probability of fuel failure due to these factors is no 
different for MOX fuel assemblies than for LEU fuel assemblies. The MOX 
fuel lead assemblies will be manufactured using the same quality 
standards that are used in the manufacture of LEU fuel, under a Quality 
Assurance program that conforms to 10 CFR 50 Appendix B. Likewise, the 
same operational procedures and precautions to preclude loose parts and 
debris in the reactor coolant will equally preclude fuel failures from 
these mechanisms for the MOX and LEU fuel assemblies.
    Other mechanisms that could potentially cause fuel cladding failure 
are physical interaction of the cladding with loose debris in the 
reactor coolant system or corrosion product transport and buildup on 
cladding material. The design of both the current LEU fuel assemblies 
and the planned MOX fuel assemblies minimizes these types of 
interactions such that the probability of fuel failure is equally 
unlikely for both MOX and LEU fuel assemblies.

Fuel Handling Accident Probability

    There is nothing in the physical design of a MOX fuel lead assembly 
that would make it more susceptible to a fuel handling accident than a 
LEU assembly. The physical dimensions are virtually identical, the 
difference in weight between a MOX assembly and a LEU assembly is less 
than 1%, and the top nozzle engages the manipulator crane and handling 
fixture in the same manner as LEU fuel.
    The shipping container and associated unloading procedure for a 
fresh MOX fuel assembly are slightly different from that of a LEU fuel 
assembly but such differences do not result in a significant increase 
in the probability of an accident. The MOX fuel lead assembly shipping 
container is an end-loaded container with capacity for one fuel 
assembly as opposed to a LEU shipping container which is side loaded 
and has the capacity for two fuel assemblies. The MOX fuel assembly 
container is unloaded by uprighting the container, removing the closure 
lid, grappling the assembly with the Fuel Handling Tool, and lifting 
the assembly with a straight vertical lift out of the container. This 
is a straightforward lifting operation that will be practiced in a dry 
run involving a dummy fuel assembly, the MOX fuel shipping package, and 
specific fuel handling procedures. The same plant equipment will be 
used to grapple and lift a MOX fuel assembly that is used to lift a LEU 
fuel assembly. Once the MOX fuel lead assemblies are unloaded and 
placed into the spent fuel pool, subsequent handling operations are 
identical to LEU fuel handling. Thus, it is concluded that the 
probability of a fuel handling accident involving a MOX fuel assembly 
drop, either inside containment or inside the fuel building, is no 
different than for a LEU assembly.
    The other scenarios considered as part of the fuel handling 
accident analyses are a weir gate drop into the spent fuel pool and a 
tornado-generated missile entering the spent fuel pool. There is no 
connection between the type of fuel assembly and the probability of 
occurrence of either of these accidents. The probability of a tornado 
missile entering the spent fuel pool is a natural event whose frequency 
of occurrence will not change with the storage of MOX fuel assemblies 
in the fuel pool. The probability of dropping a weir gate into the 
spent fuel pool is dependent on the reliability of handling fixtures, 
crane rigging procedures, and the number of handling operations, none 
of which will be affected adversely by the handling or presence of MOX 
fuel assemblies.
    The conclusion is that amending the McGuire and Catawba licenses to 
allow the receipt, handling, storage, and use of MOX fuel lead 
assemblies does not result in a significant increase in the probability 
of occurrence of any accident previously evaluated in the UFSAR.

[[Page 41854]]

NRC Staff Analysis of Consequences

    The licensee's calculated numerical values of dose consequences 
have changed since the licensee's initial submittal as addressed in the 
licensee's submittals dated November 3, 2003, March 1 and March 16, 
2004. Therefore, the NRC staff provides results from the licensee's 
submittals and the NRC staff's review that relate to an assessment of 
whether the radiological consequences from the use of MOX LTAs on 
previously analyzed design basis accident (DBA) would be expected to 
increase significantly.
    The NRC staff's review focused on the potential impacts of the 
following three characteristics of MOX fuel: (1) The fission product 
inventory in a MOX fuel assembly is expected to be different from that 
of an LEU assembly due to the replacement of uranium by plutonium as 
the fissile material, (2) the fraction of the fission product inventory 
in the gap region of a MOX fuel assembly is greater due to the 
increased fission gas release (FGR) associated with higher fuel pellet 
centerline temperatures of MOX fuel, and (3) the increased FGR can 
result in higher fuel rod pressurization.
    The configuration of the MOX LTAs is very similar to that of the 
LEU fuel assemblies currently in use at Catawba. No other plant 
modifications have been proposed by the licensee. There is no change in 
rated thermal power or any significant changes to other plant process 
parameters that are inputs to the radiological consequence analyses. As 
such, the only impacts on these analyses would be from changes in the 
fission product inventory and the gap fractions, and in the case of the 
fuel handling accident (FHA), changes in the spent fuel pool 
decontamination factor due to higher fuel rod pressurization.

Radiological Consequence Analyses

    Three categories of DBAs were analyzed for the effects of MOX LTAs.
    The first category of accidents involves damage to a significant 
portion of the entire core. They range in core damage from the locked 
rotor accident (LRA) with 11 percent core damage, the rod ejection 
accident (REA) with 50 percent core damage, to the large break loss-of-
coolant accident (LOCA) with full core damage. The results of Duke's 
analysis of these DBA categories are as follows:

    For the LRA, the four MOX LTAs represent only 19 percent of the 
21 affected assemblies in the core. The potential increase in the 
iodine release and the thyroid dose is 12 percent. The thyroid dose 
increased to 4.1 rem at the EAB, and 1.3 rem at the LPZ.
    For the REA, the four MOX LTAs represent only 4.1 percent of the 
affected 97 assemblies in the core. The potential increase in the 
iodine release and the thyroid dose is 2.63 percent. The thyroid 
dose increased to 1.03 rem at the EAB, and 0.1 rem (increase masked 
by numeric rounding) at the LPZ.
    For the LOCA, the four MOX LTAs represent only 2.1 percent of 
the 193 assemblies in the core. The potential increase in the iodine 
release and the thyroid dose is 1.32 percent. The thyroid dose 
increased to 90.2 rem at the exclusion area boundary (EAB), 25.3 rem 
at the low population zone (LPZ), and 5.37 at the control room.

    These changes in dose consequences constitute a small percent of 
the difference between the current dose value and the regulatory 
guideline value, and therefore, do not represent a significant increase 
in the consequences of these previously evaluated accidents.
    The second category of accidents includes the fuel handling 
accident (FHA), the weir gate drop accident (WGD) and the fresh MOX LTA 
drop accident. Duke assessed the MOX LTA impact on doses for the FHA 
and WGD accidents by re-calculating the analyses of record with updated 
input data. Duke projected radiological consequences to increase for 
the FHA from 1.4 to 2.3 rem Total Effective Dose Equivalent (TEDE) at 
the EAB, from 0.21 to 0.34 rem TEDE at the outer boundary of the LPZ 
and from 1.3 to 2.1 rem TEDE in the control room. Duke projected 
radiological consequences for the WGD to increase from 2.2 to 3.5 rem 
TEDE at the EAB, from 0.31 to 0.5 rem TEDE at the outer boundary of the 
LPZ and from 2.1 to 3.3 rem TEDE in the control room.
    Duke also assessed the radiological consequences of a drop of a 
fresh MOX LTA prior to it being placed in the spent fuel pool. Although 
the configuration of the MOX pellets and LTA fuel rods provides 
protection against inhalation hazards, it is conceivable that some 
plutonium might become airborne if the MOX LTA is severely damaged. The 
EAB and control room TEDE estimated by the licensee for the postulated 
fresh fuel assembly drop were less than 0.3 rem. These consequences are 
bounded by the consequences of a dropped irradiated fuel assembly.
    These resulting dose consequence values provide significant margin 
to the values specified in 10 CFR 50.67, ``Accident Source Term,'' as 
supplemented by regulatory position 4.4 of RG 1.183, ``Alternative 
Radiological Source Terms for Evaluating Design Basis Accidents at 
Nuclear Power Reactors,'' and therefore, do not represent a significant 
increase in the consequences of these accidents.
    The third category of accidents includes accidents whose source 
term assumptions are derived from reactor coolant system (RCS) 
radionuclide concentrations. These include, steam generator tube 
rupture, main steam line break, instrument line break, waste gas decay 
tank rupture, and liquid storage tank rupture. The radionuclide 
releases resulting from these events are based on established 
administrative controls that are monitored by periodic surveillance 
requirements, for example: RCS and secondary plant specific activity 
LCOs, or offsite dose calculation manual effluent controls. Increases 
in specific activities due to MOX LTAs, if any, would be limited by 
these administrative controls. Since the analyses were based upon the 
numerical values of these controls, there can be no impact of MOX LTAs 
on the previously analyzed DBAs in this category.

II. New or Different Accident Evaluation

    The proposed license amendment to allow the use of MOX fuel lead 
assemblies will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The MOX fuel assemblies have similar mechanical and thermal-
hydraulic properties to and nuclear characteristics only slightly 
different from the current LEU fuel assemblies. The use of MOX fuel 
lead assemblies does not involve any alterations to plant equipment 
or procedures that would introduce any new or unique operational 
modes or accident precursors. The existing design basis accidents 
described in the UFSAR remain appropriate and have been evaluated to 
demonstrate that there is no significant adverse safety impact 
related to the use of MOX fuel lead assemblies.
    The main physical difference between a fresh MOX fuel assembly 
and a LEU fuel assembly is the presence of more radioactivity from 
the actinides in the MOX fuel matrix, resulting in a measurable dose 
rate in the immediate vicinity of a MOX fuel assembly. As a result, 
fresh MOX fuel is transported in a sealed leaktight shipping 
container by an enclosed tractor trailer truck. There are also 
differences in the fresh MOX fuel handling procedures, but these 
differences do not lead to a new or different type of accident.
    A fuel handling accident involving a fresh MOX fuel assembly has 
potential for off-site dose consequences; however, the results of 
this fuel handling accident are bounded by the current analysis of a 
spent LEU fuel assembly drop accident. The calculated site boundary 
and control room dose consequences for a fresh MOX fuel handling 
accident are much less than the calculated doses for an accident 
involving a spent LEU fuel assembly and are well within the 
guidelines in 10 CFR Part 100. This accident does not involve a new 
release path, does not result in a new fission product barrier 
failure mode, and does not create a new sequence of events that 
would result in significant cladding failure. Therefore, this 
accident is not a new or different kind of accident.
    In conclusion, amending the * * * Catawba license to allow the 
receipt,

[[Page 41855]]

handling, storage, and use of MOX fuel lead assemblies does not 
create the possibility of a new or different kind of accident.

III. Margin of Safety Evaluation

    The proposed license amendment to allow the use of MOX fuel lead 
assemblies will not involve a significant reduction in a margin of 
safety.
    There are provisions in the * * * Catawba Technical 
Specifications that allow a ``limited number of lead test 
assemblies'' to be placed in ``nonlimiting core regions.'' These 
provisions will not change and will apply to the planned use of MOX 
fuel lead assemblies. The effect of these provisions is to place 
restrictions on the allowable power distribution limits for a MOX 
fuel lead assembly.
    The core design process assures that the limiting fuel rod in 
the core, whether LEU or MOX, has adequate nuclear power design 
limits under normal, transient, and accident conditions. If the core 
design process reveals unacceptable margin, adjustments are made to 
restore the needed margin. The operating limits are established in 
Core Operating Limits Report to assure the design limits are not 
exceeded, thus assuring that adequate design margins for the fuel 
are maintained. This iterative design process is used to analyze the 
core containing MOX fuel lead assemblies to assure that there is no 
significant reduction in a margin of safety.
    Because these lead assemblies will be located in nonlimiting 
locations i.e., will have margin above that of the limiting 
assemblies, the results of safety analyses will likewise assure that 
appropriate margins to safety are maintained during transients and 
accidents.

    On the basis of the information provided by the licensee and 
developed by the NRC staff, it appears that the three standards of 10 
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendment request involves no significant hazards 
consideration.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received. 
Should the Commission take this action, it will publish in the Federal 
Register a notice of issuance and provide for opportunity for a hearing 
after issuance. The Commission expects that the need to take this 
action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D59, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Comments on 
this Notice may also be delivered to the Commission's Public Document 
Room (PDR), located at One White Flint North, Public File Area O1 F21, 
11555 Rockville Pike (first floor), Rockville, Maryland. A copy of any 
Comments should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, and it is 
also requested that copies be transmitted either by means of facsimile 
transmission to 301-415-3725 or by e-mail to [email protected]. A 
copy of any Comments should also be sent to Ms. Lisa F. Vaughn, Legal 
Department (ECIIX), Duke Energy Corporation, 422 South Church Street, 
Charlotte, North Carolina 28201-1006, attorney for the licensee. 
Documents may be examined, and/or copied for a fee, at the NRC's PDR.
    A Notice of Consideration of Issuance of Amendment to Facility 
Operating License and Opportunity for a Hearing was published in the 
Federal Register on July 25, 2003 (68 FR 44107). On August 21 and 
August 25, 2003, respectively, the Nuclear Information and Resource 
Service and the Blue Ridge Environmental Defense League filed a 
petition requesting a hearing and seeking to intervene in the license 
amendment proceeding. Pursuant to a notice issued on September 17, 
2003, the Commission established an Atomic Safety and Licensing Board 
to preside over this matter.
    Since a hearing has been requested, the Commission will make a 
final determination on the issue of no significant hazards 
consideration. The final determination will serve to decide when the 
hearing is held. If the final determination is that the amendment 
request involves no significant hazards consideration, the Commission 
may issue the amendment and make it immediately effective, 
notwithstanding the request for a hearing. The completion of any 
ongoing hearing may take place after issuance of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    For further details with respect to this action, see the 
application for amendment dated February 27, 2003, as supplemented by 
letters dated September 15, September 23, October 1 (two letters), 
October 3 (two letters), November 3 and 4, December 10, 2003, February 
2, 2004, (two letters), March 1, 2004, (two letters), March 9, 2004, 
(two letters), March 16, 2004 (two letters), March 26, March 31, April 
13, April 16, May 13 and June 17, 2004 which are available for public 
inspection at the Commission's PDR, located at One White Flint North, 
File Public Area O1 F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (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, 301-415-4737, or by e-mail to [email protected].

    Dated at Rockville, Maryland, this 1st day of July 2004.

    For the Nuclear Regulatory Commission.
Robert E. Martin, Sr.,
Project Manager, Section 1, Project Directorate II, Division of 
Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 04-15696 Filed 7-9-04; 8:45 am]
BILLING CODE 7590-01-U