UREX + process

Uranium carbonates, 25 430-432 Uranium chlorides, 25 438-439 Uranium compounds, 25 421-434 handling, 17 529 Uranium dioxide, 25 422-423 Uranium-enrichment process gas centrifuge, 25 413-415 Uranium exploration, 25 398 URanium Extraction (UREX) process, 25 420... 

As a consequence, corporations operating PUREX plants have been using sophisticated process simulation codes, including the PAREX code in France (45-47), SpeedUp (Aspen Plus) in the UK (48), and SIMPSEX code in India (49-51). Argonne Model for Universal Solvent Extraction (AMUSE) code in the United States was contrived not only for PUREX, but for UREX+ processes (52), which will be mentioned later. In Japan, similar efforts have also been made (53-55). 

Objectives Radiotoxicity OMEGA Project3 Expecting Progress of Science and Technology in Future Decrease to 1/1000 after 100 Years of Disposal AFCI Project (UREX+1a Process)b Changeable with Variation of UREX+Processes The Requirements are not in View of Radiotoxicity but of Better Utilization of Resources ... 

Expecting Progress of Science and Technology Changeable with Variation of UREX+Processes... 

As U is the major component of a SNF see Table 1.2, its initial separation in reprocessing alleviates the mass burden of following steps and is considered preferable. The UREX process developed in the AFCI program of the United States is based on the PUREX process (30 vol % TBP in n-dodecane) and suppression of extractions of Pu and Np by reduction/complexation (175-182). Plutonium and Np are reduced by acetohydroxamic acid (AHA, CH3CONHOH) to Pu(III), Np(V), and Np(IV). U is kept in an extractable U(VI) state. Although Np(IV) is also extractable, AHA forms a complex with Np(IV) that is soluble in the aqueous phase. In the case where reoxidation of Pu(III) occurs, the Pu(IV) also transfers to the aqueous phase by forming a Pu(IV)-AHA complex. Thus, U is exclusively extracted. AHA decomposes to hydroxylamine and acetic acid (176). 

The reaction rate of AHA is large enough to use centrifugal contactors. Process experiments with real SNF in a series of centrifugal contactors have demonstrated a separation of highly pure U with a yield of >99.99% (126, 183-188). The UREX+ process also enables the effective separation of Tc (189). 

Karraker, D.G. 2002. Radiation chemistry of acetohydroxamic acid in the UREX process. WSRC-TR-2002-00283. 

Vandegrift, G.F., Regalbuto, M.C., Aase, S.B. et al. 2004. Designing and demonstration of the UREX + process using spent nuclear fuel. ATALANTE 2004, Nimes, France, June 21-24, 012-01. 

Law, J.D., Herbst, R.S., Todd, T.A. et al. 2001. The universal solvent extraction (UNEX) process. II. Flowsheet development and demonstration of the UREX process for the separation of cesium, strontium, and actinides from actual acidic radioactive waste. Solvent Extr. IonExch. 19 (1) 23-36. 

Since the launching of the Advanced Fuel Cycle Initiative (AFCI) program in the United States, the TRUEX solvent has been integrated in the live-step UREX+ process, initially consisting of live solvent-extraction steps to separate the constituents of dissolved spent nuclear fuels into seven fractions (101) ... 

The flowsheet of the UREX process, developed in the United States, includes the following extraction cycles (1) separation of uranium and technetium, (2) separation of plutonium, (3) separation of cesium and strontium, (4) separation of MAs and Rare Earth Elements (REE), and (5) group separation of MA from REE metals.9,10 Flowsheet development in Europe11 includes a modified PUREX process and, after that, the DIAMEX process for separation of MAs and lanthanides, the SANEX process for separation of MAs from lanthanides, and a special cycle for Am/Cm separation. Cesium and strontium will be in the raffinate of the DIAMEX process, and this raffinate will be vitrified, or cesium can be preliminarily extracted.12... 

PYRO-A A pyrochemical process proposed for use in nuclear reprocessing for separating transuranic elements from fission products, once the uranium has been removed by the UREX process. The spent fuel is dissolved in a molten salt bath and electrolyzed. The transuranic elements deposit on the cathode, and the fission products remain in the melt. Developed by the Argonne National Laboratory. See also PYRO-B. 

UREX process Flowsheet. (From Thompson, M.C. et al.. Demonstration of the UREX solvent extraction process with Dresden reactor fuel solutions, WSRC-TR-2002-00444, Savannah River Site, Aiken, SC, 2002.)... 

Regalbuto, M. 2011. Alternative separation and extraction UREX + processes for actinide and targeted fission product recovery. In Keimeth Nash and Greg Lumetta (eds.). Advanced techniques for nuclear fuel reprocessing and radioactive waste treatment. Chapter 7, pp. 176-200. 

There remains the question of how to come by the first core loading without separation of Pu. One possibility [XX-8, XX-33] is to use LWR spent fuel as the feed material and to remove from it only part of the uranium and part or all of the FP. For example, if the LWR spent fuel contains 1% Pu and minor activities (MA), it is necessary to remove approximately 90% of the uranium to make a fuel with 11 to 12 % of Pu and MA by weight. This could hopefully be done using a highly proliferation-resistant process, possibly a combination of an AIROX process and a fluoride volatilization process or a simplified version of the UREX process. Another feed option that could be considered is the spent fuel from MOX fuelled LWRs. The transuranium isotopes (TRU) content in such spent fuel can be approximately half of that needed for ENHS like reactors. Hence, only -50% of the uranium need be extracted along with FP to make fuel for ENHS like reactor. The latter is likely to offer a more economical fuel cycle. 

PYRO-A A pyrochemical process proposed for use in nuclear reprocessing for separating transuranic elements from fission products, once the uranium has been removed by the UREX process. The spent fuel is... 

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