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The license renewal rule amending 10 CFR Part 51 promulgated on December 18, 1996 (61 FR 66537) gave license renewal applicants the responsibility to comply with the existing requirements of 10 CFR 51.52. Section 51.52(a) specifies six conditions that must be met in order for an applicant to adopt the values in Table S-4 of that section, which represent the contribution of transportation to the environmental costs of licensing the reactor. If the six conditions are not met, an applicant must submit a full analysis of the environmental impacts of transportation of fuel and waste in accordance with §51.52(b). Two of the conditions limit the fuel enrichment level and the burnup level . Paragraph 51.52(a)(2) requires a uranium-235 enrichment not exceeding 4 percent by weight in the fuel. Paragraph 51.52(a)(3) requires that "The average level of irradiation of the irradiated fuel from the reactor does not exceed 33,000 megawatt-days per metric ton, and no irradiated fuel assembly is shipped until at least 90 days after it is discharged from the reactor." These two limiting conditions have been exceeded through nuclear power plant license amendments permitting incremental increases in the burnup of fuel. During the 1990s, the NRC has reviewed and approved vendor topical reports requesting approval for higher burnup level. (Letter from M. J. Virgilio, NRC, to N. J. Liparulo, Westinghouse Electric Corporation, "Acceptance for Referencing of Topical Report WCAP-12488, 'Westinghouse Fuel Criteria Evaluation Process," dated July 27, 1994; FCF-BAW 10186P-A, "Extended Burnup Evaluation," June 12, 1997; and Memorandum from T. E. Collins to B. W. Sheron, "Waiver of CRGR Review of EMF-85-74(P), Revision O, Supplements 1 and 2 Safety Evaluation," dated February 9, 1998). Approved average burnup for the peak rod now ranges from 50,000 to 62,000 MWd/MTU. The higher burnup levels are associated with uranium-235 enrichment levels of up to 5 percent by weight. Thus, it is likely that at the time of a submittal of a license renewal application, many nuclear power plants will be operating at higher fuel burnup and will be using higher enrichment fuel.
Further, the assumed minimum time for shipping spent fuel of 90 days after discharge from the reactor was based on the assumption that the spent fuel would be shipped to a reprocessing facility. Reprocessing spent fuel is currently not a reasonable assumption. Currently, the reasonable assumption is that spent fuel will be shipped to an interim storage facility or to an ultimate repository and would have been discharged from the reactor at least 5 years earlier and, in some cases, as many as 40 years earlier. In fact, the current practice of NRC issuing certificates of compliance for casks used for shipment of power reactor fuel is to specify 5 years as the minimum cooling period. The assumption of 5-year cooling is an extremely conservative assumption. For example, there is almost 40,000 tons of spent fuel in storage now, some of which has been stored for decades. At the earliest, if Yucca Mountain were found suitable and if DOE were successful in obtaining an NRC license, it will be at least 11 years from now until Yucca Mountain would be ready to accept spent fuel for storage. It would take many years to work off the backlog of stored spent fuel.
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Because many nuclear power plants are now operating with higher enriched fuel irradiated to higher burnup levels, motivated in part by a desire to minimize spent fuel inventory, and because of public concerns about transportation impacts of higher burnup SNF, the NRC staff examined recent technical literature on, and performed additional analyses of the characteristics of higher burnup SNF. The analyses summarized below address two questions: the extent to which higher burnup SNF might have greater incident-free transportation impacts than spent fuel with the characteristics assumed for Table S-4, whether accidents involving higher burnup SNF might have unacceptable impacts, and whether accidents involving higher burnup SNF might cause criticality during a transportation accident.
For incident-free transportation, the principal concern is whether, because of its different radiological composition, higher burnup fuel would require more shipments and larger transportation impacts than predicted by Table S-4. Quantification of the radiation emissions for reactor fuels is a complex process. However, there are several insights that allow for scaling of the radioactive emissions from one burnup level to another. For the gamma-ray sources, the scaling due to burnup is a linear relationship, i.e. a doubling of the burnup yields, a doubling of the gamma-ray emissions and, typically, a doubling of the dose rate due to gamma rays. The scaling for neutrons is not linear. Neutron emissions increase as the fourth power of the burnup ratios given the same initial enrichments; that is, doubling burnup increases the neutron emissions rate about sixteen times. In practice, however, higher burnup fuels require higher initial enrichments, such that neutron emissions typically increase as the square or cube of the burnup ratios. For example, analysis by Parks et al. (1987) showed that for a 35,000 MWd/MTU and a 60,000 MWd/MTU (burnup ratio 1.71) the neutron emissions ratio is 4.28 (less than the third power of the burnup ratio).
The increase in the total radiation dose rate due to higher burnup is complicated because the total dose rate is the sum of the gamma-ray and neutron dose rates. For nominal burnups, the dose rates at the surface and 2-m from the surface are approximately 90 percent gamma-rays and 10 percent neutrons. Indeed, Westfall et al. (1990) found that for a transportation cask that was designed for use in DOE spent fuel applications, the calculated total dose rate at 2 m for 60,000 MWd/MTU SNF was 2.19 times larger than for 35,000 MWd/MTU SNF. Thus, the total dose from a full cask of 60,000 MWd/MTU SNF would be about twice as large as the dose from a full cask of 35,000 MWd/MTU SNF. Assuming an additional increase in maximum burnup to 62,000 MWd/MTU would not invalidate that assumption given the small increase in burnup from 60,000 MWd/MTU.
The most obvious way to compensate for a doubling of the 2-m dose rate would be to halve the cask payload. This would increase the number of shipments required, but is unlikely to be pursued because of the economic and other pressures to minimize spent fuel transportation activities. In addition, under this scenario, a cask would be partially loaded (i.e. derated) with the remaining locations in the basket left empty. However, because the cask would have to be certified for higher burnup to carry even a partial load, the license submittal could easily analyze the use of inserts, which would drastically reduce the external doses with less impact on cask capacity.
There are, however, less costly ways to accommodate higher burnup fuels. Broadhead et. al. (1992) showed that by using a modified basket and by derating the cask 15 percent (an 18-assembly payload vs a 21-assembly payload) a cask with 5-year-cooled 60,000 MWd/MTU spent fuel had a lower dose rate than a 21-assembly cask containing 35,000 MWd/MTU fuel that had cooled 5 years. While the dose rates of higher-burnup fuels decline more slowly than 35,000 MWd/MTU fuel, Broadhead et al. also showed that increasing cooling times from 5 to 15 years compensates for an increase in burnup from 35,000 to 60,000 MWd/MTU. That is, a cask designed for 5-year-cooled 35,000 MWd/MTU spent fuel should be capable of accommodating 15-year-old 60,000 MWd/MTU spent fuel without derating. Thus, where on-site storage of SNF is not too costly, transportation costs and impacts can be minimized by allowing higher burnup SNF to cool 15 years before disposal.
The above two scenarios present cases where the high burnup fuels can be placed into standard casks with little or no cask derating, while meeting radiation limits outside of the cask. Under these scenarios, the actual number of trips to a repository would be decreased because the number of spent fuel assemblies required for given amount of power would be smaller with higher burnup fuel. There are other scenarios in which the number of required trips is reduced by "blending" of cask loadings, in which higher-burnup fuel assemblies are placed in the middle of the cask, while lower burnup assemblies are place near the edge of the cask cavity region to absorb radiation from the inner assemblies. While this scenario appears feasible, it has not yet been approved by NRC. A totally new cask specifically designed for high-burnup fuel is another possibility. It would only be conjecture to discuss the results of such a cask design effort, but the modified cask basket described in Broadhead et al. indicates that such a design could have little impact on the cask payload. Thus, for higher burnup fuel that is allowed to cool for at least five years before shipping, reasonable, cost-effective measures can assure that radiation limits from SNF casks can be met without increasing the number of SNF shipments. Consequently, the NRC staff concludes that use of higher burnup fuel would not lead to incident-free transportation impacts that are larger than those predicted by Table S-4.
To answer the question of how higher burnup fuel would affect accident doses, the NRC staff used the characteristics of 5 percent enriched fuel that had been burned for 62,000 MWd/MTU in its RADTRAN analysis to estimate health risks associated with accidents that release radioactive materials from a transportation cask (Table 2). The results of the analysis (Tables 3 and 4) show that higher burnup fuel has doses and health risks that are less than 15 percent of incident-free doses and health risks, and small as characterized by Table S-4.
The NRC staff also examined unlikely accident scenarios involving higher burnup fuel to determine if they could lead to a nuclear criticality event. The NRC staff examined two scenarios: failure of fuel cladding and failure of a portion of the neutron absorption material while the fuel remains in its original position. Because fuel rods are arranged in near optimum configurations for establishing and maintaining nuclear fission reactions, if the fuel cladding failed, the cask filled with water, and the fuel pellets crumbled into a pile or any other arrangement, the fuel would be farther from criticality than while they were in their original arrangement. Thus, failure of fuel cladding could not cause a criticality event.
The second hypothetical accident scenario has the neutron absorption effect at the end of the cask basket somehow lost while the fuel lattice structure remains intact. In this scenario, the fuel would remain in its optimum (for criticality) configuration, but the cask basket material which absorbs neutrons is removed from 15 cm (6 in.) of the end of the fuel rods. Analysis of several burnup levels showed that nuclear criticality would not occur, even if the cask were filled with water. Consequently, the NRC staff concludes that higher burnup SNF offers no greater criticality concerns, even in the event of unlikely occurrences.
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Most nuclear plants are now operating with higher enriched fuel irradiated to higher burnup levels than anticipated by the analyses that led to the impact levels identified in Table S-4. The NRC staff has extensively studied the environmental impacts associated with fuel enrichment up to 5 percent uranium-235 and fuel burnup to 60,000 MWd/MTU and has found that these impacts are no greater than and likely less than the impacts described in 10 CFR 51.52(c), provided that higher burnup fuel has been removed from the reactor for at least five years before it is shipped off site.
The analysis described above showed that higher enriched, higher burnup fuel would not increase incident-free-transportation or transportation-accident impacts, and that criticality could not occur during transportation of higher burnup SNF under any foreseeable circumstance. The higher burnup levels are associated with uranium-235 enrichment levels of up to 5 percent by weight. An increase in burnup from 60,000 MWd/MTU to 62,000 MWd/MTU will not significantly change dose levels associated with spent fuel transportation and may slightly reduce the number of shipments. Therefore, the impacts identified in Table S-4 bound the transportation impacts of higher enriched, higher burnup SNF. These conclusions are applicable to any nuclear power plant license renewal application provided higher burnup fuel has cooled at least 5 years before it is shipped off the reactor site.