Solvent extraction is the major unit operation employed in spent nuclear fuel reprocessing. The operation yields three streams; fission product waste, uranium product and plutonium product. Ion exchange is primarily used in reprocessing as a tail-end method to concentrate and isolate the plutonium product stream. This review will describe the details of plutonium recovery and purification by both cation- and anion-exchange processing. A brief overview of miscellaneous uses of ion-exchange employed in reprocessing will also be given.

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Grain growth behavior of UO₂ and (U, Gd)O₂ fuel pellets was investigated with the data from the out-of-pile isothermal heating experiments and the irradiation test at the Halden Boiling Water Reactor. The laboratory data gave best-fitted equations by employing the following fourth power rate equations :
where, D₀ and D are initial and final three-dimensional diameters (μm), respectively, R the gas constant (=1.987 cal/mole/K), T<> the absolute temperature (K) and t the time (h) (gadolinia content : 310W/0, temperature range : 1, 7002, 000°C).
The calculated grain diameter with the above equations revealed an overestimation on specimens which involved noticeable fission gas bubbles on their grain boundaries. It was demonstrated that the in-pile grain growth model, as was given in the following equation, which took account of the retarding effects of growth by precipitated intergranular bubbles could describe the grain growth of the irradiated samples :

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Zirconium oxide nodules formed on BWR fuel rods were characterized quantitatively and correlated statistically with the rod operational parameters. Cladding specimens were obtained from fuel rods irradiated in a commercial BWR. Their burnup and fast neutron fluence ranged 1738 GWd/t and 4×10²⁵8×10²⁵ n/m², respectively. Characteristic variables of the nodules such as maximum thickness T_max (μm) were measured on metallographs of the cladding cross sections. These variables were correlated by multiple regression analyses with the operational parameters, such as irradiation time t (d), linear heat rate p (kW/m) and fast neutron flux φ (n/m²•s). For example, the maximum thickness depended on linear heat rate and showed a saturating tendency with burnup B (GWd/t) (T_max ∝t^0.8±0.5p^2.3±0.9 or T_max∝ B^0.8±0.4p^1.5±0.5) This decrease of growth rate with irradiation time was interpreted in terms of a microstructure change of Zircaloy-2 during neutron irradiation. Results of transmission microscopy and energy dispersive X-ray spectroscopy indicated that the alloying elements such as Fe, Cr and Ni dissolved from intermetallic precipitates into the base metal during neutron irradiation. Dissolution of the alloying elements might be effective in decreasing the growth rate of nodules.

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Formation of nitrogen compounds in N₂-H₂O systems under radiation was studied. A reaction scheme with 24 species and 73 reactions was proposed. Two trigger reactions for nitrogen atom generation : N₂ ?? 2N (g-value of 1.6 μmol per 1 kJ absorbed energy in N₂ molecules), and N₂+H₂O^*→NH₂+NO (k=3.0×10⁹ _s^-1•_M^-1 at 298 K with activation energy of 34.4 kJ/mol) were included.
Calculated results with the reaction scheme agreed within an error factor of two with ammonium formation rates from an aqueous solution with dissolved nitrogen and hydrogen gasses under γ-irradiation at temperatures of 288473 K. The reaction scheme was also verified with BWR plant observations on nitric acid formation from in-leaked air under hydrogen water chemistry and ammonium injection tests.

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Photochemically induced redox reactions of Np(VI) and Np(IV) to Np(V) are experimentally studied in nitric acid solution of low Np concentration less than 10^-4 mol.dm^-3 using a Kr-F excimer laser. The preparations of the initial solution of Np(VI) or Np(IV) are based on the addition of chemical redox reagents. The extraction chromatography is used to analyze the fraction of the neptunium valences. The result of redox experiments with laser radiation shows that the fractions of Np(V)/Np(VI) or Np(V)/Np(IV) photochemically increase up to the steady-state values against initially oxidizing or reducing conditions respectively. The steady-state values are different from those at thermal equilibrium states. It is concluded that, in low Np concentration as is observed in the normal Purex process, laser application to valence adjustment of Np gives another redox condition which is different from that determined thermochemically.

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The depletion and production amounts of U, Pu, transplutonium nuclides and fission products (FPs) measured on the fuel of JPDR-1 were corrected to take account of the performance history of irradiation and cooling using the results of three-dimensional nuclear-thermo-hydrodynamic and nuclide depletion and production calculations. Except a few nuclides, the corrected values proved to agree well with the values calculated by the ORIGEN computer code. Further enhancement of calculational accuracy calls for systematic re-evaluation of neutron cross sections on the basis of neutron spectrum in nuclear fuel.

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Fuel behavior during a reactivity initiated accident condition is recognized to be predo minantly related to energy deposition in the fuel. The first stage of NSRR in-pile experiments addressing the behavior of PuO₂-UO₂ mixed oxide fuels evaluated the energy deposition per unit integrated reactor power by γ-ray spectrometry. Solid samples were used to measure the γ-rays because the facility is permitted to handle solid plutonium only. Determination of the penetration ratios of γ-rays from the actinides contained in the fuels allowed correction for the self-attenuation of γ-rays in the solid samples. Evaluation of the effect of epithermal neutron fissions was also necessary since the fissile nuclides of ²⁴¹Pu and ²³⁹Pu have high resonance cross sections in the epithermal energy region. For this evaluation, the fission density was first calculated for the fission products as a function of the contribution ratio of the fissions of epithermal neutrons. Accurate fission density was then determined using the contribution ratio which minimized the deviation of the calculated values for the fission density. The fission densities determined by this simplified method agreed well with the values calculated using the computer codes CITATION and GGC-4.

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The migration prediction system (MIGSTEM) has been developed for estimating the migration of cationic species of radionuclides through soil layers systematically. The MIGSTEM consists of the migration experiments, the one-dimensional fitting code (inverse analysis code) for determining retardation factor and dispersivity (migration factors) and the three-dimensional differential code (prediction code) for estimating the migration of the radionuclides. The migration experiments are carried out for obtaining the concentration profiles of the radionuclides in unsaturated and saturated soil layers. Using the inverse analysis code, the migration factors are obtained at one time by fitting the concentration profiles calculated to those observed. The prediction code can give the contours of concentration and the one-dimensional concentration profiles at selected time, as well as the changing in the concentration at a selected position with time. The validity of the MIGSTEM was obtained by the benchmark test on the pre diction and inverse analysis codes. The MIGSTEM was applied to estimate the migration of Sr²⁺ through the sandy soil.

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An electromagnetic (EM) flow coupler works as a sodium pump. It consists of two parallel ducts under a transverse magnetic field. Conducting fluid in one of the ducts is driven by external power (a mechanical or an electromagnetic pump). Then the duct works as a MHD generator and current is supplied to the other duct. Under the transverse magnetic field, the induced current provides a pumping force to the fluid in the adjacent duct, realizing a high efficiency sodium pump. When the EM flow coupler is used for FBRs, the secondary sodium flow can be used as the generator flow. Then the primary sodium flow is obtained only by the duct under a transverse magnetic field. In the case of an annular EM flow coupler, the ducts are assigned in the annular channel. The annular channel is divided in the circumferential direction, and the generators and pumps are arranged alternatively. Since a radial magnetic field is applied in the flow ducts, the induced current circulates around the flow ducts. So, no bus bar, using a material such as copper which cannot be placed in a high temperature sodium environment, is required. Further, the coupler's compact arrangement around the reactor core can decrease the size of the reactor pool.
Experimental results of an annular prototype which was made of stainless steel and had no insulator (non-insulated prototype) have been discussed based on its equivalent electric circuit analysis⁽¹⁾. For the non-insulated prototype, the maximum efficiency was less than 30% since the ohmic loss in the side wall was more than 40% at the peak efficiency. So, an efficiency of more than 60% would be expected if electrical insulation of the side wall were done. In this note, the experimental results are given about the effect that the side wall insulator has on the efficiency.

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A cryogenic freezer system is one of the most useful systems in connection with the hydrogen isotope recovery, which is expected to be used in a fuel cycle of a D-T fusion reactor, particularly in a fuel gas refining system. As an example of the cryogenic freezer system, it is designed to freeze water vapor containing tritium at a temperature of 160 K from all condensable impurities eluted from a molecular sieve bed⁽¹⁾⁽²⁾. It is very important to investigate the influence of the variables on a mass flux at cooling walls of the freezer. The effects of frost formation on cooling walls and mist formation in a gas stream on an average mass flux in a cooling pipe were experimentally and theoretically studied and were presented in a separate paper⁽³⁾. It is considered that there are cases where foreign nuclei of impurities are present in the cryogenic freezer, because the gas eluted from the molecular sieve bed is generally composed of multi-component condensable mixtures. In this note the basic experiments and the numerical calculations were carried out for the investigation of the effects of the presence of foreign nuclei in a gas stream on the mass flux. The NaCI nuclei which are generated by the evaporation and the fast cooling-down of a NaCl solid in an electric furnace⁽⁴⁾ were used in this experiment as the foreign nuclei.

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One of the most important factors for measuring precise leach rates in continuous-flow leach tests on high-level waste glass is the dilution of leachate with freshly pumped leachant⁽¹⁾⁽³⁾. When analyzing results from the MCC-4 leach tests⁽⁴⁾, the factor of dilution by leachant flow should be taken into account, since the factor determines the apparent saturation concentration after a long duration ; the higher the flow rate, the lower the apparent saturation concentration becomes⁽³⁾. The dilution of leachate with freshly pumped leachant is examined here, and it is pointed out that the diffusion of elements in water should be fast enough to achieve uniform mixing in leach containers.

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