Neptunium from nuclear fuel waste

Nagasaki, S., Wisnubroto, D.S., Enokida, Y., Suzuki, A. 1994. Neptunium separation from nitric acid solutions with CMPO. 4th Int. Conf. Nuclear Fuel Reprocessing and Waste Management, RECOD 94 Proceedings Vol. 3, London, UK, April 24-28. 

Each of these elements may be used for production of nuclear fuel or other purposes. The recovery efficiency for uranium is reported as 99.87% and for plutonium 99.36%-99.51% (NEA 2012). The extended PUREX includes separation of neptunium and technetium as well as recovery of americium and curium that are also separated from each other by additional extraction stages as given in detail in the flowsheet (NEA 2012). The advanced UREX-i-3 process generates six streams after separation uranium for re-enrichment Pu-U-Np for mixed oxide fuel c for managed disposal Am-Cm to be used as burnable poisons and for transmutation high-heat-generating products (Cs and Sr) and a composite vitrified waste with all other fission products. Some fuel types may require preliminary steps like grinding to enable their dissolution. 

Neptunium-237 Np-237 has two production routes. The first is rapid /3 decay of U-237 produced by nuclear reactions in the fuel. The second route is /3 decay of Pu-241 to Am-241, which decays losing an a particle to Np-237. This second route will lead to the in-growth of Np-237 in the waste during storage. Np-237 will only be present in waste contaminated with actinides. This is demonstrated by the waste streams that were found to exceed the GQ level. They were IX resin and sludge from BWA and SPF eontents from HPA. Adequate estimates of Np-237 can be made by ratio to Cs-137 determined by FISPIN in the first instance and then refined by taking account of the accumulation and chemistry of other transuranics already measured. At higher concentrations, it can also be measured by y spectrometry of the short-... 

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