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Projected Salt Waste Production from a Commercial Pyroprocessing Facility

DOI: 10.1155/2013/945858

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Abstract:

Pyroprocessing of used nuclear fuel inevitably produces salt waste from electrorefining and/or oxide reduction unit operations. Various process design characteristics can affect the actual mass of such waste produced. This paper examines both oxide and metal fuel treatment, estimates the amount of salt waste generated, and assesses potential benefit of process options to mitigate the generation of salt waste. For reference purposes, a facility is considered in which 100?MT/year of fuel is processed. Salt waste estimates range from 8 to 20?MT/year from considering numerous scenarios. It appears that some benefit may be derived from advanced processes for separating fission products from molten salt waste, but the degree of improvement is limited. Waste form production is also considered but appears to be economically unfavorable. Direct disposal of salt into a salt basin type repository is found to be the most promising with respect to minimizing the impact of waste generation on the economic feasibility and sustainability of pyroprocessing. 1. Introduction Pyroprocessing is currently being developed by several nations (including but not limited to the USA, Republic of Korea, Russia, and India) for potential implementation in a large-scale, commercial fuel cycle. In most cases, this is driven by the process’ deemed compatibility with a fast reactor. There is also a nonproliferation element to its attractiveness. Japan is the only nuclear nonproliferation Treaty (NPT) nonnuclear weapons state that possesses full-scale nuclear fuel reprocessing facilities. They use a PUREX-based process to coextract uranium and plutonium from their spent fuel. Due to the potential to adapt this process to extract weapons-grade plutonium, Japan’s use of the technology has come under widespread international criticism. In an effort to avoid such controversy, the Republic of Korea has been developing pyroprocessing since 1997 for the purpose of closing the fuel cycle and managing their used fuel stockpiles [1]. Pyroprocessing has been viewed as being significantly less versatile with respect to producing weapons-grade materials. Another attractive feature is that the process can be designed into compact facilities colocated on reactor sites to minimize the risk of transporting used nuclear fuel on public transportation routes. In the Republic of Korea (South Korea), pyroprocessing is the only used fuel management technology option currently being considered because of the diplomatic situation on the Korean peninsula. The pyroprocessing flowsheet is conceptually simple with a

References

[1]  K. C. Song, H. Lee, J. M. Hur, J. G. Kim, D. H. Ahn, and Y. Z. Cho, “Status of pyroprocessing technology development in Korea,” Nuclear Engineering and Technology, vol. 42, no. 2, pp. 131–144, 2010.
[2]  J. P. Ackerman, T. R. Johnson, L. S. H. Chow, E. L. Carls, W. H. Hannum, and J. J. Laidler, “Treatment of wastes in the ifr fuel cycle,” Progress in Nuclear Energy, vol. 31, no. 1-2, pp. 141–154, 1997.
[3]  L. J. Simpson and D. J. Wronkiewicz, “Evaluation of standard durability tests towards the qualification process for the glass-zeolite ceramic waste form,” in Proceedings of the 20th Scientific Basis for Nuclear Waste Management, W. Gray and K. Knecht, Eds., vol. 465, pp. 441–448, Materials Research Society, 1997.
[4]  C. Pereira, “Production of sodalite waste forms by addition of glass,” Ceramic Transactions, vol. 61, p. 389, 1997.
[5]  S. Priebe and K. Bateman, “The ceramic waste form process at Idaho National Laboratory,” Nuclear Technology, vol. 162, no. 2, pp. 199–207, 2008.
[6]  M. F. Simpson, K. M. Goff, S. G. Johnson et al., “A description of the ceramic waste form production process from the demonstration phase of the electrometallurgical treatment of EBR-II spent fuel,” Nuclear Technology, vol. 134, no. 3, pp. 263–277, 2001.
[7]  E. J. Karell, K. V. Gourishankar, J. L. Smith, L. S. Chow, and L. Redey, “Separation of actinides from LWR spent fuel using molten-salt-based electrochemical processes,” Nuclear Technology, vol. 136, no. 3, pp. 342–353, 2001.
[8]  S. D. Herrmann, S. X. Li, M. F. Simpson, and S. Phongikaroon, “Electrolytic reduction of spent nuclear oxide fuel as part of an integral process to separate and recover actinides from fission products,” Separation Science and Technology, vol. 41, no. 10, pp. 1965–1983, 2006.
[9]  S. X. Li, S. D. Herrmann, and M. F. Simpson, “Electrochemical analysis of actinides and rare earth constituents in liquid cadmium cathode product from spent fuel electrorefining,” Nuclear Technology, vol. 171, no. 3, pp. 292–299, 2010.
[10]  S. X. Li, S. D. Herrmann, K. M. Goff, M. F. Simpson, and R. W. Benedict, “Actinide recovery experiments with bench-scale liquid cadmium cathode in real fission product-laden molten salt,” Nuclear Technology, vol. 165, no. 2, pp. 190–199, 2009.
[11]  M. F. Simpson, T. S. Yoo, D. LaBrier, M. Lineberry, M. Shaltry, and S. Phongikaroon, “Selective reduction of active metal chlorides from molten LiCl-KCl using lithium drawdown,” Nuclear Engineering and Technology, vol. 44, no. 7, pp. 767–772, 2012.
[12]  C. Pereira, M. C. Hash, M. A. Lewis, M. K. Richmann, and J. Basco, “Incorporation of radionuclides from the electrometallurgical treatment of spent fuel into a ceramic waste form,” Materials Research Society Symposium Proceedings, vol. 556, p. 115, 1999.
[13]  Y. Z. Cho, H. C. Yang, H. C. Eun, E. H. Kim, and I. T. Kim, “Characteristics of oxidation reaction of rare-earth chlorides for precipitation in LiCl-KCl molten salt by oxygen sparging,” Journal of Nuclear Science and Technology, vol. 43, no. 10, pp. 1280–1286, 2006.
[14]  V. A. Volkovich, T. R. Griffiths, and R. C. Thied, “Treatment of molten salt wastes by phosphate precipitation: removal of fission product elements after pyrochemical reprocessing of spent nuclear fuels in chloride melts,” Journal of Nuclear Materials, vol. 323, no. 1, pp. 49–56, 2003.
[15]  Y. Z. Cho, G. H. O. Park, H. S. U. Lee, I. N. T. Kim, and D. S. Han, “Concentration of cesium and strontium elements involved in a LiCl waste salt by a melt crystallization process,” Nuclear Technology, vol. 171, no. 3, pp. 325–334, 2010.
[16]  B. J. Riley, B. T. Rieck, J. S. McCloy, J. V. Crum, S. K. Sundaram, and J. D. Vienna, “Tellurite glass as a waste form for mixed alkali-chloride waste streams: candidate materials selection and initial testing,” Journal of Nuclear Materials, vol. 242, no. 1–3, pp. 29–37, 2012.
[17]  Y. Wang, M. Simpson, J. Rath et al., “Closing the nuclear fuel cycle with salt,” in Proceedings of the 13th International High-Level Radioactive Waste Management Conference, Albuquerque, NM, USA, 2011.

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