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An Introduction to the Purex Process

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An Introduction to the Purex Process An Introduction to the Purex Process . IntroduCtion allowing for the decay of short-lived fission products is termed as spent fuel reprocessing. During the operation of nuclear reactors, 235u component of the natural uranium fuel is fissioned, Process Requirements generating neutrons and a host of highly radioactive fission products. Simultaneously, the fertile 238u Fuel reprocessing differs from conventional captures some of the neutrons and yields 239Pu, anew chemical processing due to the radioactive nature of. and useful fissile nuclide. the materials being processed. The equipment have to be installed behind massive concrete shielding 2 ~W + /,n ~ Fission products + 2.5 ~n + 194 MeV (sometimes as much as 1.5 M .thick) to provide built-in arrangement for separation and protection of the operating personnel from equipment with high 238U I 239lJ ~- 239 R- 239n radiation field. Stringent air ventilation and exhaust 92 + on ~ n : > 93Np ~d 94t"U 23 .5 mtn. 2 .3 ay requirements are to be met in the entire plant to The accumulated fission products in the fuel protect the operating personnel and the environment hinder the operation of the reactor system from from getting contaminated due to air borne radiation and neutron economy points of view. radioactivity. The processes and equipment used are modified to suit remote operation and maintenance. Becau~e of this, as the burn up (expressed as MWD/tonne) of the fuel goes high, the spent fuel has Another important difference between to be replaced with fresh fuel in the reactor to traditional and nuclear chemical engineering is the . continue its operation. ' need to provide a design that precludes the possibility The spent fuel discharged from reactors of accidentally producing a self-sustaining nuclear contains significant quantities of fissile nuclides, chain reaction- the condition known as criticality­ 235 when lar~ concentrations or quantities of fissile mainly the unutilized U and the newly formed 39 235 233 239Pu which can be re-used. The process of isotopes ( Pu, U or U) are handled. Safe separating the fissile and fe1tile materials (Pu and U) operation of a reprocessing plant is generally from the . fission products in the spent fuel after achieved by criticality control techniques like mass, volume, and concentration control of fissile materials or geometry control of the equipment used. July 1998 lJ IANCAS Bulletin . msr crm I~ o, ,'.....' co-mncn. "'"o1 uol LUI 0 SOLV. POl SCRUB I P UCTCLI SOL.I . G c 0 L > L 10. rm O,Pa,PP . L l Q.O PIODUCr ) soLb uu lOt rBP/DD SOLY. FOI UCYCLI rr v1sn roa com .< c i STOilGI 0 ·-, - -------------L'<LllCI CYCLIO VAST£ _jVE:TI . AQ.Pa PRODUCt PP, 0 ' PI lRd fa tYCU Fig. 1 Flow scheme of Purex process Purex~ • Final U purification cycle • Final Pu purification cycle World over, Purex process is followed for the • Pu and U reconversion to their oxides I I separation and purification of uranium (U) and plutonium (Pu) from irradiated spent fuels after The first three steps together are called dissolving the fuels in nitric acid. Purex process is head-end treatment steps meant to bring the fuel into essentially a solvent extraction process which uses dissolved aqueous form suitable for Purex process. Tri-n-butylphosphate (TBP) diluted with dodecane The aim of the co-decontamination and partitioning as solvent to extract and further separate U and Pu step is to extract and separate U and Pu together from from highly radioactive fission products. The U and the rest of the fission product impurities and then Pu products are further purified and converted to further separate and purify U and Pu from each other. oxides, while fission product · waste solution is In the subsequent steps, U and Pu are purified further neutralized and stored for subsequent treatment and individually before conversion to their oxide forms. disposal. The general flow scheme used is given in Fig. I . The extent of purification or decontamination of U or Pu achieved in each step is generally The Purex process has the following major expressed by a term called decontamination factor steps: (DF) and is defined by the following expression: • Decladding or dejacketing p, y acti vity/g ofU or Pu in feed • Dissolution DF = rl . I f U P . d • Feed preparation p , y acttvtty go or u 111 pro uct • Co-decontamination and Partition cycle 1ANCAS Bulletin 12 July 1998 The larger the DF one obtains in an extraction fission products by sparging oxygen as reactant and step, the greater is the purification and efficiency of this type of dissolution is called fumeless dissolution. the process. In general, the DF with respect to beta and gamma activities aimed in Purex process after 2 Feed Preparation 6 7 cycles of extraction is of the order of 10 to 10 for The solution resulting from the dissolution of both uranium and plutonium. spent fuels must be adjusted in most cases before it is ready for solvent extraction involving one or more Head End Process Steps in Purex Process of the following operations: Dec/adding • Solids removal or feed clarification to avoid Fuels have different cladding materials such as choking, pluggin~ and emulsification during aluminium, zircaloy and stainless steel depending on extraction the reactor types: The removal of jacket material • Adjustment of feed acidity to obtain the prior ·co processing is desirable, as otherwise it extraction of the required component contributes greatly to the volume of highly • Adjustment of salting strength by dilution or radioactive process solution to be processed and the evaporation of the feed to the desired uranium waste that must be stored. Based on the cladding content level, to have maximum processing rate, material, the decladding method is selected, ie. wherever necessary. chemical decladding or mechanical decladding. In the former method, the decladding material is In Purex process, at the end of the dissolution, preferentially dissolved in aqueous solution whereas uranium is present in sixth valence state and the fuel core· remains undissolved. In the latter plutonium mostly in tetravalent state. For good method, the fuel assemblies are chopped and extraction by TBP, tetravalent state of total dissolved in nitric acid and the clad· material plutonium is ensured by further addition of NaN02 remaining undissolved is disposed as solid waste. followed by digestion at 50°C. This technique is mostly adopted for spent fuel 4 arisings from power reactors (PHWRs), as they are Puoi+ + N02 + 2H; ~ Pu + + N03 + H20 ge.nerally clad with zircaloy. Pu3+ + N02 + 2H+ ~ Pu4+ +NO+ H20 Dissolution 6Pu3+ + 2N02 + 8 H+ ~ 6Pu4+ + N2 + 4 H20 Concentrated nitric acid is used to dissolve uranium and uranium oxide fuels. Dissolution is The dual function nitrite (i.e. it reduces Pu(VI) carried out batch-wise. The dissolution is exothermic to Pu(IV) and oxidizes Pu(Ill) to Pu(IV)) makes it an and the reaction is controlled by optimizing the ideal feed conditioning agent in Purex process. N02 concentration of acid as well as the temperature of gas can also be used for this adjustment, as it reaction. Oxides of nitrogen liberated during the precludes the introduction of sodium ion as impurity dissolution are reoxidised and put back in the and has other advantages as well. dissolver using down draft condensers to reduce the. Solvent Extraction with TBP consul)lption of nitric acid. Fm1her, the gases are treated to remove the traces of nitlic acid and volatile As the major separation process steps are fission products, cooled and filtered before being essentially solvent extraction cycles with TBP, the exhausted through a tall stack. ~p extraction technique will be discussed briefly before -~ discus~ion on these process steps is taken Uranium dioxide dissolves in nitric acid by the · up. net reaction: TBP as an Exiractant Several factors influence the choice of a solvent It is possible to dissolve uranium without any for the 'e~tt~ction process in the processing of net evolution of gaseous products except the gaseous irradiated fuel elements. A final choice involves a July 1998 13 IANCAS Bulletin compromise . between various factors. TBP is in the aqueous phase form nitrate complexes that are · 4 generally favoured as it meets most of these extracted byTBP. ForUO~+, Puo~+, Pu + and for few conditions and is employed in most of the fuel other ions in te.travalent state, complexing is reprocessing plants in the world. TBP is highly considerable. But for a few exceptions like Ru(NO), 4 selective for U and Pu. As a commercial product, it Zr +, etc., the ability to form nitrate complexes is is cheaply available and can be purified easily. It has very poor for most of the elements that constitute the a high boiling point (266°C) and is non-volatile. Its fission products. solubility in water is_ Qnly 0.4 g/.!:jt has lflgJ1" chenucai;-Ihermal and radiation sta5ility. It has very Because of the similarity of U(VI) and Pu(IV) low extractability for fission products resulting in in the formation and extractability of nitrate excellent purification of U and Pu. The only two complexes and the. inextractability of fission unfavourable properties are its high density (0.973) products, an efficient separation of fanner from the and high viscosity which are compensated by latter is obtained by solvent extraction with TBP. diluting it with an inert diluent like kerosene. In the Then taking advantage of the fact that Pu(III) is process, it can be purified by washing with weakly extracted, a further separation ofU and Puis NaOH/Na2C03 and HN03 to remove the achieved by reducing Pu(IV) to Pu(III), in which decomposition products. The difficulties in the state it gets stripped from the organic phase.
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