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 Start Printed Page 41853and 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. Start Printed 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, Start Printed Page 41855handling, 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 OGCMailCenter@nrc.gov. 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 firstname.lastname@example.org.Start Signature
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