The effect of ionizing radiation caused either by photons or particles having sufficient energy to ionize the molecules of the medium is very important in process chemistry. It may involve photons with energies ranging from the first ionization energy of the medium (~10 eV) up to several MeV, as well as alpha and beta particles (electrons) generated by spent nuclear fuel decay. Whether in solid, solution, or gaseous states, the alpha, beta and gamma radiation interacts with the environment and affects the chemical speciation of the actinides because the result of the energy absorption is breaking or rearrangement of chemical bonds.

A portion of ATM-109 fuel was removed from its cladding, dissolved in concentrated nitric acid, filtered, and partially evaporated. The equal ali­quots of the dissolved fuel solution were diluted to the same volume by the variable nitric acid solutions to adjust uranium concentration to approxi­mately 0.7 M and the acid strength to 0.3, 1.3, 2.5, 3.8, and 5.1 M to mimic the feed stream for a PUREX or UREX process in uranium concentration and provide a range of acid variations that might occur in the feed stream. The uranium and other fuel component concentrations remained the same in each sample, regardless of the acid strength.

Using a batch technique for the first stage of the PUREX process, the acid-adjusted aqueous solutions were contacted with equal volume of a 30 vol% TBP/n-dodecane solvent. The mixtures were allowed to reach an

4.10 Raman spectra of (A) ATM-109 BWR commercial fuel 0.3 M HNO3 feed, raffinate, and simple feed simulant solution (1.33 M uranyl nitrate in 0.8 M HNO3); and (B) TBP-dodecane extraction solutions of ATM-109 feed and simple feed simulant solutions.

equilibrium distribution; then, the phases were separated. The feed, raffinate, and extract samples were subjected to spectroscopic analysis as described in the next section.

The COEXTM or COEXtraction process has been developed by French workers, and is designed to bleed some of the U (and possibly Np) into the Pu product to eliminate the pure Pu stream (Baron 2007). The development of this new reprocessing and recycling COEXTM process was aimed at:

• further enhancing proliferation resistance

• maintaining a high level of process performance

• minimizing both investment (capital) and operational costs

• taking full advantage of present industrial experience

• keeping open possible evolutions to take into account new types of reactors or future changes in management strategies of the transuranic elements.

The separation flowsheets used in the COEXTM process (Drain 2008) are based on the expertise accruing from the design, and deployment of the PUREX process, but equally from operational feedback with that process. The design for the extraction cycles has extensively relied on a simulation tool, validated by experimental investigations carried out in the workshops at the La Hague plants, and equally by comparisons with findings from industrial operations. Consequently, the COEXTM process is basically an

evolution of the PUREX process, by modifying it to produce a U + Pu mixture (U/Pu >20%), rather than pure plutonium: no pure plutonium is separated at any point of the process (see Fig. 6.7). The advantage of so doing is to curb proliferation risks and to produce a perfectly homogeneous mixed oxide, for the purposes of MOX fuel fabrication, affording enhanced performance.

The proposed COEXTM flowsheet is indicated in Fig. 6.7, and involves an initial U/Pu codecontamination step that is fairly similar to the first purifica­tion cycle, as implemented in the La Hague and Rokkashomura plants (see Fig. 6.2). The codecontamination step is unchanged: this covers the extrac­tion flowsheet, including measures to allow removal of technetium. The plutonium stripping part is modified by including conditions that results in uranium scrubbing into the Pu strip product. This function is slightly tweaked from uranium scrub to neptunium scrub, to allow some uranium to remain in the plutonium stream. The function further allows neptunium to be extracted from the plutonium stream, directing the neptunium thus extracted to the cycle’s uranium product stream, thanks to higher distribu­tion coefficient of Np(IV) compared to U(IV) (see Table 6.1).

To ensure the plutonium becomes extractable by the solvent phase again, an adjustment is made to the solution yielded by the first cycle, by raising the nitric acid concentration, concomitant with reoxidation of plutonium to oxidation state IV, uranium being adjusted to oxidation state VI. The stream then undergoes treatment in the form of a U-Pu cycle, only involving
plutonium extraction, and back-extraction (stripping) functions. Reductive plutonium stripping is effected using hydroxylamine nitrate.

Topping up with uranium is provided for, at the head end of this opera­tion, to adjust the Pu/U ratio in the production stream. Finally, the stripped solvent is recycled in the plutonium stripping operation carried out upstream, in the first cycle. This measure makes it possible to tolerate some loss of plutonium to the stripped solvent, and thus limit the number of stages used for the plutonium stripping operation (hence the term “mini­cycle,” for this complementary purification cycle).

A uranium purification cycle, identical to the cycles implemented at the La Hague plant, is further provided for — on the one hand, to complete decontamination of the uranium stream, with respect to в and у emitters, and, on the other hand, to ensure removal of neptunium.

A de facto separation and concentration of fission products can be achieved by crystallization of bulk fission-product components, such as occurs during nitric acid recovery evaporations. In advanced reprocessing schemes, such as the Japanese NEXT process,56 this approach takes the form of an initial crystallization of uranyl nitrate. For cleanup of Hanford alkaline tank waste, proposed fractional crystallization of sodium nitrate achieves an equivalent result.57,58 Given that low concentrations of strontium are present in alkaline wastes, especially those with complexants, a useful decontamination can be achieved by isotopic dilution upon addition of strontium nitrate and result­ing precipitation of strontium carbonate.33 In the mid 1990s, the ITP process for cesium removal by precipitation with sodium tetraphenylborate was implemented at SRS.59 Owing to the insolubility of cesium tetraphenylbo — rate, an impressive decontamination of 1 Ci/L alkaline waste can be reduced to <10 pCi/L. However, due to the action of trace catalysts to accelerate tetraphenylborate decomposition, unexpectedly large benzene levels occurred in the tank headspace and led to shutdown of this process. A modification of the process, called the Small Tank Precipitation Process, was developed to overcome the benzene evolution by exploiting the induction period of the catalyst. The modified process was successfully demonstrated60 but has not been adopted for use.

Although the raffinate salt contains very low concentrations of actinides, it still needs to be treated at the ‘zeolite column’ process to remove LFP ele­ments, by ion exchange with structural elements of zeolite or by occlusion of chloride molecules in the 3-dimensional cage structures of zeolite. Figure 10.7 shows the equilibrium absorption characteristics of alkali metal chlo­rides, alkaline-earth chlorides and lanthanide chlorides in zeolite-4A at temperatures ranging from 400 °C to 450 °C (Tsukada, 2008). As the figure shows, the amount of absorption depends on the fraction of the element in the salt and on the valence of the element. At the same equivalent fraction of approximately 0.1, the amount of lanthanide elements loaded is about twice that of alkali metal elements, while the amount of alkaline-earth ele­ments loaded is about 1.5 times that of alkali metal elements. Though further study is still needed to gain detailed absorption behaviours, it is clear that zeolite-4A works as a potential absorbent of LFP from LiCl-KCl eutectic salt. To use zeolite as an LFP absorbent, a continuous system with a zeolite column is preferable from the point of view of efficiency; however, a batch-type system with a zeolite bed is also possible. Whichever system is employed, the LFP-loaded zeolite still needs further treatment to form

10.7

Absorption characteristics of alkali metal chloride, alkaline-earth chloride and lanthanide chloride in zeolite 4A.

acceptable waste because a certain amount of free salt containing LFP adheres to the surface of the zeolite.

In France and in Europe, the bidentate oxygen-donor solvation extractants that were investigated in the 1990s to co-extract An(III) and Ln(III) were the N, N,N’,N’-tetraalkyl-malonamides (Gasparini and Grossi, 1980, 1986, Musikas, 1987, 1988, Thiollet and Musikas, 1989, Cuillerdier et al., 1991,1993, Nigond et al., 1994a, b, 1995, Madic and Hudson, 1998): RR’N(C=O)-CHR"- (C=O)NRR’, a subgroup of the diamide family bearing a methylene bridge linking the two amide functions (R, R’, and R" representing linear or branched alkyl or phenyl substituents). Malonamides enable the formation of energetically stabilized 6-membered ring complexes when they extract trivalent metallic cations.

Malonamides were designed to compete with carbamoyl-phosphine oxide compounds, such as CMPO (Fig. 11.8), developed in the 1980s at the Argonne National Laboratory (ANL, USA) to decontaminate transuranic waste (Navratil and Thomson, 1979, Horwitz et al., 1981, 1982) arising from the production of military grade plutonium. Although of lower efficiency than CMPO when extracting An(III) from nitric acid solutions, lipophilic

malonamides (containing a suitable number of carbon atoms) present several advantages over CMPO:

• Malonamides are soluble in hydrogenated tetrapropene: concentrations exceeding 1 mol. L-1 can be dissolved in HTP. Therefore, malonamides do not require any phase modifier to cope with the relatively high solvent loading capacities required by PUREX raffinates.

• As malonamides present steeper nitric acid dependence than CMPO, the stripping of extracted elements is facilitated.

• Malonamides are less stable than equivalent neutral organophosphorus compounds, especially versus acidic hydrolysis, but their carboxylic deg­radation products are less detrimental to the back-extraction of the minor actinides in diluted nitric acid.

• As they are made of carbon, hydrogen, oxygen, and nitrogen atoms, malonamides generate only mineral ashes after incineration.

The affinity of malonamides toward An(III) and Ln(III) decreases as the atomic number of the extracted element increases. The stoichiometry of the extracted complexed M(III) cation is assumed to be ML2(NO3)3 (where L = malonamide) at saturation, although higher stoichiometries have been observed owing to malonamide aggregation. The use of small angle neutron/ X-ray scattering techniques and the application of colloidal concepts to malonamide solvents actually proved that these compounds self-organize in small aggregates (Fig. 11.9), consisting of spherical polar cores, mainly

composed of the polar heads of 4 to 10 malonamide molecules and of the extracted solutes (0.5 < фсот<, < 1.2 nm), surrounded by non polar crowns, mainly composed of the alkyl chains of the malonamides and of the diluent (Abecassis et al., 2003, Bauduin et al., 2007, Dozol and Berthon, 2007).

It must be remembered, though, that aggregation can easily become detrimental to process implementation, because the polar core attractions can induce the splitting of the loaded solvent into two phases: an enriched one in metallic complexes and another, lighter (called ‘third phase’), usually composed of almost pure diluent. The stability of an organic phase contain­ing aggregates depends on the equilibrium between the different interac­tions: (i) the attractive van der Waals forces taking place between the aggregate polar cores, (ii) the repulsive forces between the polar cores (assimilated as hard, sticky balls in the Baxter model), and (iii) the repulsive forces between the aggregates, due to the stabilizing repulsive steric interac­tions between the hydrophobic alkyl chains of the extractant and those of the diluent.

Due to the amphiphilic nature of the malonamides, the corresponding solvents behave like reverse micro-emulsions stabilized by surface active compounds. The attractive forces depend on the composition of the polar cores: they increase with the concentration of the extracted solute and depend on its nature. The aggregate repulsive forces depend on the length of the alkyl chains of both the malonamide extractant and the diluent mol­ecules, which act in opposite manner: long alkyl chains for the malonamide and short alkyl chains for the diluent prevent third phase formation, which will occur as soon as the attractive forces between the aggregates exceed their repulsive forces (Berthon et al., 2007).

Development of the DIAMEX process

The structure of the DIAMEX process reference extractant has evolved between 1991 and 2001, from N, N’-DiMethyl-N, N’-DiButyl-TetraDecyl — MalonAmide (DMDBTDMA, Fig. 11.8, Musikas et al., 1991) to N, N’- DiMethyl-N, N’-Octyl-HexylEthoxy-MalonAmide (DMDOHEMA, Fig. 11.8, Madic et al., 2002), in order to:

• increase the total number of carbon atoms, to enhance both the hydro — phobicity of the extractant and the solubility of the extracted metallic complexes in HTP, and hence prevent third phase formation;

• improve the extractant affinity toward trivalent metallic cations, by introducing an ethoxy moiety in the alkyl chain grafted onto the meth­ylene bridge (Spjuth et al., 2000);

• uniformly display the carbon atoms among the alkyl chains grafted onto the malonamide to simplify the elimination of its degradation com­pounds coming from acidic hydrolysis and radiolysis (Berthon et al., 2001): basic washings of the degraded solvent have proved to be efficient in getting rid of the hydrophilic acidic degradation compounds.

The DMDOHEMA flowsheet was directly adapted from that of DMDBTDMA thanks to the PAREX process simulator code, developed at the CEA. Several counter-current tests have been carried out from 1999 to 2005, both at the CEA Marcoule (France), at Forschungszentrum Julich (FZJ, Germany), and at the Institute for Transuranium Elements (ITU, Karlsruhe, Germany) during successive collaborative projects funded by EURATOM (Courson et al., 2000, Madic et al., 2000, 2002, 2004, Christiansen et al., 2004, Warin, 2007). These tests put into operation: [12]

pulsed columns (two for the extraction of An(III) and Ln(III), and one for FP scrubbing), eight mixer-settlers (for the stripping of An(III) and Ln(III)), and four centrifuges (for the spent solvent clean-up) is shown in Fig. 11.10 (Warin, 2007). This test, as well as a long-term hydrolysis/radiolysis endur­ance test, in which the DMDOHEMA solvent was recycled after specific caustic washings to eliminate its acidic degradation compounds (such as carboxylic acids and acid-amide), validated the industrial applicability of the DIAMEX process by demonstrating the possible recovery of more than 99.9% of An(III) and Ln(III) from a genuine highly active PUREX raffi­nate, with high decontamination factors, DF, toward fission products (e. g., DFZr ~ 800).

Currently available treatment options utilize mostly physical or chemical processes to separate the toxic component of the waste for further treat­ment. This is done to achieve one or all of the four targets for handling of waste, i. e. waste minimization, toxicity reduction, volume reduction, and/or

Porous

carbon

buffer

15.2 Propagation of fission products and impurities in a graphite regulated high temperature (gas) cooled reactor fuel element.

security (deterrence of proliferation). These targets can be achieved through physical, chemical, or biological means described below.

K. L. NASH and J. C. BRALEY, Washington State University, USA

Abstract: From the days of the Manhattan Project, the chemistry of actinides and selected fission products has shaped decisions on the handling of irradiated nuclear fuel. This chemistry is characterized by the diversity of the fission products, the rich redox chemistry of the light actinides, high radiation levels, concentrated nitric acid used to dissolve the fuel and the nuclear chemistry of both actinides and fission product lanthanides. This chapter introduces the actinide and fission product chemistry relevant to the nuclear fuel cycle, from the isolation of uranium from mined ores through reprocessing to management of the byproduct wastes. The important features of historically successful solvent extraction separations and alternative chemical processes are described. Finally, the role of nuclear energy as a source of primary power sans greenhouse gases is discussed.

Key words: nuclear fuel, actinide chemistry, solvent extraction, molten salts.

1.1 Introduction

Prior to the 1940s, the only radioactive elements on planet earth were those long-lived enough to have persisted since the condensation of the solar system (primarily 235, 238U, 232Th, 40K, 87Rb), the short-lived isotopes linking actinides to their stable end-member lead or bismuth isotopes (Ra, Rn, Po…), those created as a result of cosmic radiation in the upper atmosphere (36Cl, 32P, 14C, 3H…) and the very small amounts of man-made isotopes that had been created in scientific research. That research and the activities of the ensuing Manhattan Project introduced to the terrestrial environment considerable amounts of both short — and long-lived new isotopes, including most significantly kilogram quantities of some transuranium elements. These activities also required the mining and processing of uranium ores, which increased the accessibility of uranium and its radioactive daughters to the biosphere. The subsequent activities of the Cold War increased the terrestrial abundance of some of these elements to thousands of kilograms. During the years of atmospheric testing of nuclear weapons, weapons tests released significant amounts of radioactive debris into the environment.

Since the institution of a ban on atmospheric testing of nuclear weapons, the only significant injection of anthropogenic radioactivity into the ter­restrial environment occurred in the Chernobyl accident in 1986.

Though nuclear energy was first exploited for plutonium production for military purposes (and was in fact driven by the defense buildup of the Cold War), the production of electricity through the operation of fission reactors began in the early 1960s. The first nuclear reactors designed for electricity production were of a graphite-moderated, gas-cooled design in the United Kingdom. In the US, the first power reactors were designed for use in US Naval ships. The same design ultimately was adapted to stationary applica­tions in water cooled and moderated reactors fueled with partially enriched uranium. The Canadian design utilized natural uranium or slightly enriched fuel with D2O moderation and cooling. Each design has its advantages and limitations, but all produce a similar array of waste byproducts, hence offer similar constraints on the execution of the fuel cycle.

The use of high purity light water (H2O) as a neutron moderator and coolant offers several advantages, starting with its reasonable price and favorable thermal properties. In the operation of a light water reactor, enriched uranium (3-5% 235U, 97-95% 238U) is partially consumed with the resulting creation of transuranium elements (in order of decreasing amounts, plutonium (Pu), neptunium (Np), americium (Am) and curium (Cm)), and fission products including varying amounts of all elements in the periodic table between zinc and erbium. The fission products include noble gases, halides, calcogenides, pnictides, geranium, tin, indium, second row transition metals, alkali metals, alkaline earths and about two-thirds of the lanthanide series. Post removal of the fuel from the reactor, the total uranium content is about 95.5% of the non-oxide mass. Plutonium isotopes account for about 0.9% of the heavy metal content. The isotopic distributions for uranium and plutonium post-irradiation are shown in Fig. 1.1. Weapons grade plutonium is defined as material containing at least 93% 239Pu. Reactor grade pluto­nium is defined as material composed of more than 18% 240Pu. The high

Plutonium 238

239

240

24% sssss 241 ssss 242

neutron capture cross-section of 240Pu aids in limiting the utility of pluto­nium with significant amounts 240Pu from being used in a weapon. At dis­charge, used fuel also includes about 500 g/ton of 237Np.

Though the mass of used fuel is predominantly uranium, the radioactivity of the actinide component is dominated on discharge by the plutonium and curium isotopes, the former based on mass, the latter on the short half-lives of the isotopes present. Most of these isotopes have longer half-lives than the majority of fission products, hence they tend to dominate the radiotoxic­ity of used fuel beyond about 500 years after discharge. In storage, the isotopic distribution of actinides in used fuel changes primarily from the decay of 242Cm, 244Cm, 241Pu resulting in an increase in the 238Pu, 240Pu, and 241Am content of the fuel, respectively. Though it represents only a minor component of used fuel on discharge, 241Am content of the used fuel increases substantially as a result of 241Pu decay; the dominant curium iso­topes decay away in a relatively short time (resulting in a decrease in the total curies). The total radioactivity arising from the actinide isotopes is about 0.11 MCi/ton at discharge and 0.07 MCi/ton after ten years of decay storage (these figures and all subsequent product information are based on 33,000 MWd/t U burnup at a power density of 30 MW/t U and neutron flux of 2.92*1013 Ncm-2s-1, as described in [1]). As all actinide isotopes are radio­active, ultimately all decrease in concentration of spent fuel in storage, though clearly the long-lived isotopes decrease slowly. A summation of the radioactive properties of the important actinide isotopes present in used fuel is shown in Table 1.1.

Among the fission products, there are many short-lived and stable species. At discharge, the radioactive materials content of one ton of fuel is about 15 MCi, dominated by small amounts of very short-lived fission products. The most important fission product radionuclides in used fuel after about one year are 137Cs (1230 g/ton of 30.1 year half-life in equilibrium with its radioactive 137mBa daughter — 0.21 MCi total), 90Sr (543 g/ton of 28.5 year half life in equilibrium with its radioactive 90Y daughter — 0.15 MCi total), 99Tc (841 g/ton of 2.1 x 105 year half-life — 14 Ci total) and 129I (229 g/ton of 1.57 x 107 year half-life — 0.04 Ci total). The latter two isotopes are of greater concern for their potential environmental mobility, their long half-lives and for their bioaccumulation possibilities than the numbers of curies present or the energetic characteristics of their emissions. The cesium/barium and strontium/yttrium isotopes dominate the dose, produce substantial amounts of heat and are the most radiotoxic materials in used fuel from shortly after discharge through several hundred years.

As noted above, rare earth elements are significant byproducts of fission. They represent about 40% of the mass of fission products and including measurable amounts of all lanthanides from lanthanum (La) through erbium (Er) plus moderate amounts of yttrium (Y), which exhibits chemistry similar

to that of the lanthanides. The lanthanide composition of fuel irradiated as described above (33,000 MWd/t U at power density of 30 MW/t U and neutron flux of 2.92*1013 Ncm-2s-1) is shown in Fig. 1.2. [1] This is an impor­tant factor in the used fuel management equation, as several isotopes of these elements have thermal neutron capture cross-sections greater than those of the actinides, hence must be separated to enable any waste man­agement scenario that includes transmutation of actinides by neutron capture. Figure 1.2 also contains the thermal neutron capture cross-sections and resonance integrals for important lanthanides and americium. In a light water reactor, transmutation (by fission) of Am isotopes will be virtually impossible in the presence of lanthanide isotopes. As will be noted below, the separation of fission product lanthanides from actinides can be a chal­lenging obstacle to effective waste management.

As is true of all electricity production technologies based on fuel con­sumption, nuclear power systems operate within the framework of a fuel cycle that includes mining and preparation of the fuel, “combustion” with the generation of heat to power generators, and waste management. In a nuclear fuel cycle, uranium is mined, converted and isotopically altered before fuel elements are prepared, assembled into a critical mass and allowed to undergo controlled nuclear fission. Ultimately, the fuel must be replaced with a fresh load of fuel. Unlike power generation based on fossil fuels, the large majority of “waste” byproducts in nuclear fuel are retained within the fuel assemblies in a fission reactor.

Materials considered as waste in used fuel may simply be disposed of as waste (an open fuel cycle), or recycled to recover fuel for reuse and to improve waste management (a closed fuel cycle). Both waste management methods require the accepted use of a geological repository engineered to retain the majority of the most radiotoxic elements for an adequate time to protect the surrounding environment. An open fuel cycle focuses on permanent disposal of the fuel without concern about the additional energy

potential of the irradiated fuel. The closed fuel cycle focuses on recovering additional fuel value through either recycle of plutonium or recovery of fuel value and transmutation of troublesome isotopes. Both cost and com­plexity increase as additional processing is imposed, but with the advantage of extending the potential life of fuel supplies and a shorter time require­ment for geologic isolation. The decision is one whose dimensions are defined by the need for power, supplies of resources, and the limits of safety.

In the following discussion, the basic chemistry that supports the deci­sion-making process for these options will be discussed. The emphasis will be on the solution phase chemistry of actinides and important fission prod­ucts, the former including uranium, thorium, neptunium, plutonium, ameri­cium and curium; the latter, cesium, strontium, iodine, technetium and the lanthanides. The interrelated features of the nuclear and radiochemistry, oxidation-reduction chemistry, solution chemistry (i. e., complexation and hydrolysis) media will be discussed.

Rotary vacuum filters are typically used to separate precipitated high value products such as plutonium oxalate from the “mother liquor”. An example design of rotary vacuum filter, used routinely over many years at a number of reprocessing plants, is shown in Fig. 3.5.

In this application, plutonium nitrate solution is fed to the precipitator vessels and oxalic acid used to precipitate plutonium oxalate. The resulting slurry is fed via nozzles onto a filter table which is formed of a stainless steel fine mesh and which is rotated at a few rpm. The underside of this mesh table has a vacuum applied to it which draws the mother liquor away from the precipitated solids. As the table rotates the accumulated solids move under a wash station where they are washed with water, and then under a “scraper” system that removes them from the table and collects them in a screw feeder for transfer to the next process stage. By suitably balancing the slurry feed rate and the rotation speed of the table, continuous operation can be achieved. The unit is small and compact enough to be installed within a glovebox for radioactive contamination containment, but it is not suitable for in-cell use so is not used for materials emitting penetrating radiation.

R. S. HERBST, Idaho National Laboratory, USA, P. BARON, CEA, France and M. NILSSON, University of California Irvine, USA

Abstract: The PUREX process for separating uranium and plutonium from irradiated nuclear fuels has been extensively studied and successfully operated industrially over the preceding five plus decades. It is anticipated that PUREX will play an important role in upcoming and advanced future nuclear fuel cycles. The first objective of this chapter is to provide the background information required to status the present state of the art as currently practiced at the industrial scale. The second objective is to examine the modifications ready, or nearly so, for implementation into the next generation of PUREX reprocessing facilities, thereby further expanding the utility and operation of the process for use in nuclear fuel cycles of the future.

Key words: PUREX, tributyl phosphate, uranium separation, plutonium separation, solvent extraction, nuclear separations, nuclear fuel cycle.

5.2 Introduction

The PUREX (Plutonium, Uranium, Reduction, EXtraction) process was first reported in 1949 and subsequently operated with irradiated nuclear fuel at the large scale in 1954 (Lanham 1949). In the ensuring 50+ years, it has been widely described, publicized, and reviewed in the open literature, and remains the only technology internationally practiced on an industrial scale to reprocess (recover U and Pu) from used nuclear fuels. A major objective of this chapter is then, not to provide an in depth review of PUREX technology (that has already been eloquently accomplished in a plethora of previous reports, articles, and treatises), but rather to provide the necessary background information to logically status the current state of the art, as practiced today, at the industrial scale. The second, and perhaps more timely objective is to provide insight into the modifications ready or near-ready to be implemented into the next generation of PUREX repro­cessing facilities, thereby further expanding the utility and operation of the process in the ever-changing climate of technical, political, and environmen­tal considerations.

It should be noted that initially, PUREX was developed and applied primarily to produce a pure Pu stream for military (weapons) applications and was consequently cloaked in great secrecy under the auspices of “national security.” Those countries that have produced substantial amounts of Pu for military purposes have done so by means of the PUREX process. Today, the Pu and U products are commercially recovered from used fuels from civilian nuclear power plants (NPP) and fabricated into fresh fuel for recycle back to NPPs for electricity production. Interestingly, many of the intricate details associated with present-day commercial reprocessing facili­ties are considered “propriety information” and are therefore often pro­tected as well as (or better than) what were formerly considered to be national secrets.