Standard Gibbs energy of formation of Cs2CdI4

The vapor pressures of CdI₂ and Cs₂CdI₄ were measured below and above their melting points, employing the transpiration technique. The standard Gibbs energy of formation Δ_fG° of Cs₂CdI₄, derived from the partial pressure of CdI₂ in the vapor phase above and below the melting point of the compound could be represented by the equations Δ_fG°Cs₂CdI₄ (±6.7) kJ mol⁻¹=−1026.9+0.270 T (643 K≤T≤693 K) and Δ_fG°{Cs₂CdI₄} (±6.6) kJ mol⁻¹=−1001.8+0.233 T (713 K≤T≤749 K) respectively. The enthalpy of fusion of the title compound derived from these equations was found to be 25.1±10.0 kJ mol⁻¹ compared to 36.7 kJ mol⁻¹ reported in the literature from differential scanning calorimetry (DSC). The standard enthalpy of formation Δ_fH°_298.15 for Cs₂CdI₄ evaluated from these measurements was found to be −918.0±11.7 kJ mol⁻¹, in good agreement with the values −920.3±1.4 and −917.7±1.5 kJ mol⁻¹ reported in the literature from two independent calorimetric studies.     

Post-irradiation examination of uranium-doped rock-like oxide fuels

Post-irradiation examinations of rock-like oxide fuels were performed in JAERI to evaluate irradiation behavior and geochemical stability. Five kinds of fuels were prepared using 20% enriched U instead of Pu; a single-phase fuel of an yttria-stabilized zirconia containing UO₂ (U-YSZ), two particle-dispersed type fuels of U-YSZ particles + MgAl₂O₄/Al₂O₃ powder, two homogeneously blended type fuels of U-YSZ powder + MgAl₂O₄/Al₂O₃ powder. The fuels were irradiated in JRR-3 for about 100 days and estimated irradiation conditions were as follows; linear power was 15 kW m⁻¹, thermal neutron fluence was 7 x 10²⁴ m⁻² and fuel temperatures at the surface were 740–1130 K. From the results of non-destructive examinations, the stainless steel cladding surfaces were partially discolored by oxidation and no remarkable deformation of the pins was observed. Significant pellet fragmentation was not observed in spite of the crack formation as observed in irradiated LWR UO₂ fuels. Nonvolatile FPs were observed only in pellets but volatile Cs moved partly to the plenum. From these examinations, no significant difference in macroscopic irradiation behavior was distinguished among 5 fuels.     

Rock-like oxide fuel behavior under reactivity initiated accident conditions

Pulse irradiation tests of two types of rock-like oxide (ROX) fuel, i.e. yttria stabilized zirconia (YSZ) and YSZ/Spinel composite, were conducted in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior under reactivity-initiated accident conditions. The ROX fuels failed with cladding burst at fuel volumetric enthalpies above 10 GJ m⁻³, which was comparable to that of UO₂ fuel. The failure of the ROX fuels, however, occurred with considerable fuel melting and was quite different to that of UO₂ fuel, which was caused by cladding melting and embrittlement due to heavy oxidation. Lower fuel melting temperature of the ROX fuels compared to that of UO₂ contributed to the different fuel failure modes. Certain amount of molten ROX fuel dispersed out at the failure. However, the mechanical energy generation due to the molten fuel/water interaction was negligible for the ROX fuels at peak fuel enthalpies below 12 GJ m⁻³.     

Radiotoxicity hazard of U-free PuO2+ZrO2 and PuO2+ThO2 spent fuels in LWR

The radiotoxicity hazard of U-free Rock-like oxide: ROX (PuO₂+ZrO₂) and Thorium oxide: TOX (PuO₂+ThO₂) LWR spent fuels is investigated and radiotoxicity hazard of MOX spent fuel is considered as a reference case. The long-term ingestion radiotoxicity hazard of ROX spent fuel is one third and nearly one fourth of that of TOX and MOX spent fuels, respectively. This is because the discharged Pu and long lived Np in ROX fuel is less than that of TOX and MOX fuels. In TOX fuel, discharged Pu and MA are lower than that of MOX fuel but the long-term radiotoxicity hazard of spent fuel is nearly the same as MOX spent fuel. At the cooling 10⁵ years, the radiotoxicity hazard of TOX spent fuel is approximately ten and three times higher than that of ROX and MOX spent fuels, respectively due to higher toxic contribution of ²²⁹Th in TOX spent fuel.     

Application of lattice constant measurements for stoichiometry determination in irradiated U-Pu mixed oxide fuels

A method is proposed for determining the oxygen/ metal ratio in mixed irradiated uranium-plutonium oxides. The method is based on a measurement of the lattice constants and on a standard thermal treatment which is used to obtain a tetravalent state of uranium and plutonium. The effects of irradiation and of the solution of solid fission products in the matrix, the variation in plutonium concentration, and influence of these factors on stoichiometry are discussed on the basis of the results of simulated experiments in which the state after irradiation of oxide fuels is computed, together with the concentrations of the fission products.

For a given burn-up τ the oxygen/metal ratio of the matrix O/U + Pu + FP, which has a considerable influence on the physical properties of the fuel, is obtained by direct measurement of the ratio O/U + Pu and correcting this value for the effect of soluble fission products using the equation: O/U + Pu + F.P. = 3/(3− τ) [(1− τ)O/U + Pu] + [2/3 τ·1.75].     

Separation of plutonium from lanthanum by electrolysis in LiCl–KCl onto molten bismuth electrode

This work presents a study on the electroseparation of plutonium from lanthanum using molten bismuth electrodes in LiCl–KCl eutectic at 733 K. The reduction potentials of Pu³⁺ and La³⁺ ions were measured on a Bi thin film electrode using cyclic voltammetry (CV). A difference between the peak potentials for the formation of PuBi₂ and LaBi₂ of approximately 100 mV was found. Separation tests were then carried out using different current densities and salt phase compositions between a plutonium rod anode and an unstirred molten Bi cathode in order to evaluate the efficiency of an electrolytic separation process. At a current density of 12 mA/cm²/wt% (Pu³⁺), only Pu³⁺ ions are reduced into the molten Bi electrode, leaving La³⁺ ions in the salt melt. Similar results were found at two different Pu/La concentration ratios ([Pu]/[La] = 4 and 10). At a current density of 26 mA/cm²/wt% (Pu³⁺), co-reduction of Pu and La was observed as expected by the large negative potential of the Bi cathode during the separation test.     

Standard molar Gibbs free energy of formation of PbO(s) over a wide temperature range from EMF measurements

The EMF of the following galvanic cells,

(render)

Kanthal,Re,Pb,PbOCSZO₂ (1 atm.),Pt

(render)

Kanthal,Re,Pb,PbOCSZO₂(1 atm.),RuO₂,Pt

were measured as a function of temperature. With O₂ (1 atm.), RuO₂ as the reference electrode, measurements were possible at low temperatures close to the melting point of Pb. Standard Gibbs energy of formation, Δ_fG⁰_mβ-PbO was calculated from the emf measurements made over a wide range of temperature (612–1111 K) and is given by the expression: Δ_fG⁰_mβ-PbO±0.10 kJ=−218.98+0.09963T. A third law treatment of the data yielded a value of −218.08 ± 0.07 kJ mol⁻¹ for the enthalpy of formation of PbO(s) at 298.15 K, Δ_fH⁰_mβ-PbO which is in excellent agreement with second law estimate of −218.07 ± 0.07 kJ mol⁻¹.     

Utilization of rock-like oxide fuel in the phase-out scenario

Utilization of rock-like oxide (ROX) fuel in light water reactors for plutonium (Pu) burning was studied by nuclear material balance (NMB) analysis for a case of Japanese phase-out scenario under investigation after the Fukushima accident. For the analysis, the NMB code was developed with features of accurate burn-up calculation, flexible combination of reactors and fuels, and an ability to estimate waste and repository. Three scenario groups of once-through Pu burning in mixed oxide (MOX) fuel and in ROX fuel were analyzed. Using two full-MOX or full-ROX reactors the Pu amount is reduced to about one-half and the isotopic vector of Pu deteriorated for being used as a nuclear weapon, especially in terms of spontaneous fission neutron generation. Effects of ROX reactors are more significant than MOX reactors in terms of both reduction in the Pu amount and deterioration of the isotopic vector. Repository footprint and potential radiotoxicity are not reduced by the MOX and ROX reactors because the heat and toxicity of MOX and ROX spent fuels are considerably high.     

The closed on-site fuel cycle of the brest reactors

The BREST fast reactor with nitride fuel and lead coolant is being developed as a reactor of new generation, which has to meet a set of requirements placed upon innovative reactors, namely efficient use of fuel resources, nuclear, radiation and environmental safety, proliferation resistance, radwaste treatment and economic efficiency. Mixed uranium-plutonium mononitride fuel composition allows supporting in BREST reactor CBR≈1. It is not required to separate plutonium to produce “fresh” fuel. Coarse recovered fuel purification of fission products is allowed (residual content of FPs may be in the range of 10⁻² – 10⁻³ of their content in the irradiated fuel). High activity of the regenerated fuel caused by minor actinides is a radiation barrier against fuel thefts. The fuel cycle of the BREST-type reactors “burns” uranium-238, which must be added to the fuel during reprocessing. Plutonium is not extracted during reprocessing being a part of fuel composition, thus exhibiting an important nonproliferation feature.

The radiation equivalence between natural uranium consumed by the BREST NPP closed system and long-lived high-level radwaste is provided by actinides (U, Pu, Am) transmutation in the fuel and long-lived products (I, Tc) transmutation in the blanket. The high-level waste must be stored for approximately 200 years to reduce its activity by the factor of about 1000.

The design of the building and the entire set of the fuel cycle equipment has been completed for the demonstration BREST-OD-300 reactor, which includes all main features of the BREST-type reactor on-site closed fuel cycle.     

Structure and kinetic studies of U(VI)-benzamidoxime complex in non-aqueous solutions by H- and C-NMR

The structural and kinetic studies of U(VI) complex with benzamidoxime(Hba) as ligand in CD₃COCD₃ have been studied by means of ¹H and ¹³C NMR. The Hba molecule was found to coordinate to UO₂²⁺ in the form of anionic benzamidoximate (ba), and the number of ba coordinated to UO₂²⁺ was determined to be 3 by analyzing the chemical shift of ¹³C NMR signal for Hba in the presence of UO₂²⁺. The exchange rate constants(k_ex) of ba in [UO₂(ba)₃] were determined by the NMR line-broadening method. The kinetic parameters were obtained as follows: k_ex(25°C) = 3.1 × 10³s⁻¹, ΔH^≠ = 35.8 ± 3.5 KJ mol⁻¹, and ΔS^≠ = −65 ± 13.7 J K⁻¹ mol⁻¹. The UV-visible absorption spectra of solutions containing UO₂²⁺ and Hba were also measured. The molar extinction coefficient of the complex was found to be extremely large compared with those of UO₂(L)₅²⁺ (L = unidentate oxygen donor ligands) complexes. This is due to the strong electron withdrawing of UO₂²⁺ from Hba and suggests that an interaction between UO₂²⁺ and Hba is very strong. Such a high affinity of monomeric amidoxime to UO₂²⁺ reasonably explains the high adsorptibility of amidoxime resin to U(VI) species, and is considered to result in the high recovery of U(VI) species from sea water using amidoxime resin.     

Measurement and calculation of neutron densities in BWR high burn-up fuels

A neutron-scanning device was developed for measuring accurate neutron densities of BWR high burn-up fuels up to 65 GWd tU⁻¹. Characteristic test of this device was done with a ²⁵²Cf source and adopted to measure axial distributions of neutron densities of BWR spent fuels with various enrichments (2.0–3.4%), which had been irradiated up to 60 GWd tU⁻¹ at Fukushima Daini Nuclear Power Station Unit 2(2F-2). We found the measured neutron densities were proportional to about fourth power of the corresponding burn-up values. The neutron densities calculated by the ORIGEN2.1 code and various cross section libraries showed good agreements with the measured ones in profile and absolute value except for BWR-UE file mainly based on ENDF/B-IV. The BS240J32 library based on JENDL3.2 was the best among the investigated libraries.     

Thermodynamic studies on Cs4U5O17(s) and Cs2U2O7(s) by emf and calorimetric measurements

The oxygen potentials over the phase field: Cs₄U₅O₁₇(s)+Cs₂U₂O₇(s)+Cs₂U₄O₁₂(s) was determined by measuring the emf values between 1048 and 1206 K using a solid oxide electrolyte galvanic cell. The oxygen potential existing over the phase field for a given temperature can be represented by: Δμ(O₂) (kJ/mol) (±0.5)=−272.0+0.207T (K). The differential thermal analysis showed that Cs₄U₅O₁₇(s) is stable in air up to 1273 K. The molar Gibbs energy formation of Cs₄U₅O₁₇(s) was calculated from the above oxygen potentials and can be given by, Δ_fG⁰ (kJ/mol)±6=−7729+1.681T (K). The enthalpy measurements on Cs₄U₅O₁₇(s) and Cs₂U₂O₇(s) were carried out from 368.3 to 905 K and 430 to 852 K respectively, using a high temperature Calvet calorimeter. The enthalpy increments, (H⁰_T−H⁰₂₉₈), in J/mol for Cs₄U₅O₁₇(s) and Cs₂U₂O₇(s) can be represented by, H⁰_T−H⁰_298.15 (Cs₄U₅O₁₇) kJ/mol±0.9=−188.221+0.518T (K)+0.433×10⁻³T² (K)−2.052×10⁻⁵T³ (K) (368 to 905 K) and H⁰_T−H⁰_298.15 (Cs₂U₂O₇) kJ/mol±0.5=−164.210+0.390T (K)+0.104×10⁻⁴T² (K)+0.140×10⁵(1/T (K)) (411 to 860 K). The thermal properties of Cs₄U₅O₁₇(s) and Cs₂U₂O₇(s) were derived from the experimental values. The enthalpy of formation of (Cs₄U₅O₁₇, s) at 298.15 K was calculated by the second law method and is: Δ_fH⁰_298.15=−7645.0±4.2 kJ/mol.     

Radiotoxicity hazard of inert matrix fuels after burning minor actinides in light water reactors

Ingestion radiotoxicity hazard index of inert matrix spent fuels are investigated after burning minor actinide (MA) isotopes in LWRs and compared with the hazard index of MOX and MA burning MOX (MOX+MA) spent fuels. As U-free fuels, ROX: (PuO₂+ZrO₂) and TOX: (PuO₂+ThO₂), are considered, in which MA's are added as oxides. The radiotoxicity hazard index of ROX+MA spent fuel is less than that of TOX+MA and MOX+MA spent fuels due to the lower density of actinides in spent fuel. Some of cooling years the toxic yield of ROX+MA spent fuel is even less than that of MOX spent fuel, if the initial loaded MA in ROX is about 0.5 at %.     

Diffusion analysis on the thermal release of tritium from a neutron absorbing material

Tritium released from neutron irradiated borosilicate glass was determined by a specially designed sampling system and a liquid scintillation counter at temperatures in the range of 200–700°C. It was found that the chemical form of tritium released was tritiated water (HTO, T₂O) for the most part. Tritium produced in the glass would react with oxygen to form OT and diffuse out by a similar mechanism as the molecular diffusion of water in glasses. The diffusion coefficient of tritiated water in borosilicate glass obtained is expressed by D (cm²/s) = 5.3 × 10⁻⁴ exp( −128 kJ/mol)/RT). It is concluded from the diffusion analysis that the greater part of tritium produced in a neutron absorber, which is made of borosilicate glass, would remain in the glass for a few years of irradiation.     

甲基膦酸二甲庚酯对硝酸介质中Pu(Ⅳ)的萃取

研究了HNO₃介质中甲基膦酸二甲庚酯(DMHMP)对Pu(Ⅳ)的萃取性能,考察了DMHMP浓度、NO^-₃浓度、HNO₃浓度以及温度对Pu(Ⅳ)分配比的影响。确定了DMHMP萃取Pu(Ⅳ)的萃合物的组成为Pu(NO₃)₄·2DMHMP,其萃取反应方程式为:■其中Pu(Ⅳ)与NO^-₃形成中性分子,再与DMHMP结合成为中性配合物进入有机相。在实验范围内Pu(Ⅳ)分配比与DMHMP浓度的平方、NO^-₃浓度的四次方成正比,萃取过程为放热反应,反应的焓变为-34.46 kJ/mol。     

Fuel performance evaluation of rock-like oxide fuels

The concept of the rock-like oxide (ROX) fuel has been developed for the annihilation of excess plutonium in light water reactors. Irradiation tests and post-irradiation examinations were carried out on candidate ROX fuels. The ternary fuel of YSZ–spinel–corundum system, the single-phase fuels of YSZ, the particle-dispersed fuels of YSZ in spinel or corundum matrix, and the blended fuels of YSZ and spinel or corundum matrix were fabricated and submitted to irradiation testings. The fuels containing spinel showed chemical instabilities with the vaporization of MgO component, which caused fuel restructuring. The swelling behavior was improved with the particle-dispersed fuels. However, the particle-dispersed fuels showed higher fractional gas release (FGR) than blended type fuels. The FGR of YSZ single-phase fuels were comparable to what would be expected for UO₂ fuel at the similar fuel temperatures. The YSZ single-phase fuel showed the best irradiation performance among the ROX fuels investigated.     

A measurement of the irradiation-induced creep of mixed carbide nuclear fuel

In the BR 2 reacior at Mol, Belgium, a measurement of the irradiation induced creep of mixed carbide nuclear fuel up to high burnup was carried out The dependence upon applied stress and burnup of 95% dense (U, Pu) C was measured within a temperature range between 500 and 720°C and at fission rates between 1.0−1.5 × 10¹⁴ f/cm³ s. The used irradiation device was a Confluent-type capsule that allowed a variation of stress as well as temperature during irradiation. The length changes of the fuel specimen were determined by means of the microwave cavity resonance method. The obtained creep rates are proportional to stress and burnup-independent. The irradiation creep rates are about one order of magnitude below those of mixed oxide fuel. The fission product swelling rate increased with burnup form initially 1.2 to 3.0 vol% per % burnup. At stress changes the fuel showed a transient swelling up to 0.2 vol%. The theoretical background of carbide irradiation creep is briefly discussed.     

Irradiation effects of swift heavy ions in actinide oxides and actinide nitrides: Structure and optical properties

Actinide oxides have been used as nuclear fuels in the majority of power reactors working in the world and actinide nitrides are under investigation for the fuels of the future fast neutron fission reactors developed in Forum Generation IV. Radiation damage in actinide oxides UO₂, (U_0.92Ce_0.08)O₂, and actinide nitride UN has been characterized after irradiation with swift heavy ions. Fluences up to 3 × 10¹³ ions/cm² of heavy ions (Kr 740 Mev, Cd 1 GeV) available at the CIRIL/GANIL facility were used to simulate irradiation in reactors by fission products and by neutrons. The macroscopic effects of irradiation remains very weak compared with those seen in other ceramic oxides irradiated in the same conditions: practically no swelling can be measured and no change in colour can be observed on the irradiated part of a polished face of sintered disks. The point defects in irradiated actinide compounds have been characterized by optical absorption spectroscopy in the UV–Vis–NIR wavelength range. The absorption spectra before and after irradiation are compared, and unexpected stability of optical properties during irradiation is shown. This result confirms the low rate of formation of point defects in actinide oxides and actinide nitrides under irradiation. Actinide oxides and nitrides studied are >40% ionic, and oxidation state of the actinides seems to be stable during irradiation. The small amount of point defects produced by radiation (<10¹⁶ cm⁻²) has been identified from differences between the absorption spectrum before irradiation and the one after irradiation: point defects in oxygen or nitrogen lattices can be observed respectively in oxides and nitrides (F centres), and small amounts of U⁵⁺ would be present in all compounds.     

Formation of uranium mononitride by the reaction of uranium dioxide with carbon in ammonia and a mixture of hydrogen and nitrogen: II. Reaction rates

Kinetics of the carbothermic synthesis of UN from UO₂ in an NH₃ stream and a mixed 75% H₂ + 25% N₂ stream were studied in the temperature range of 1400–1600°C by X-ray analysis and weight change measurement of the sample. The weight change was divided into two parts; i.e. weight loss due to carbothermic reduction of UO₂ and weight loss due to removal of carbon by hydrogen. The former followed the first-order rate equation −1n(1 − ₀) = k₀t, and the latter the rate equation of phase boundary reaction 1 − (1 − _c)^1/3 = k_ct. The apparent activation energy of the former was in the range of 320–380 kJ/mol. The value of the latter in an NH₃ stream was 175–185 kJ/mol, which was smaller than that in a mixed 75% H₂ + 25% N₂stream (285 kJ/mol). In this method, the rate of the removal of carbon by hydrogen determines that of the formation of high purity UN.     

Surface morphology of the Ar ion-irradiated Ag/Si(1 1 1) system

In the present study, a 500 Å thin Ag film was deposited by thermal evaporation on 5% HF etched Si(1 1 1) substrate at a chamber pressure of 8×10⁻⁶ mbar. The films were irradiated with 100 keV Ar⁺ ions at room temperature (RT) and at elevated temperatures to a fluence of 1×10¹⁶ cm⁻² at a flux of 5.55×10¹² ions/cm²/s. Surface morphology of the Ar ion-irradiated Ag/Si(1 1 1) system was investigated using scanning electron microscopy (SEM). A percolation network pattern was observed when the film was irradiated at 200°C and 400°C. The fractal dimension of the percolated pattern was higher in the sample irradiated at 400°C compared to the one irradiated at 200°C. The percolation network is still observed in the film thermally annealed at 600°C with and without prior ion irradiation. The fractal dimension of the percolated pattern in the sample annealed at 600°C was lower than in the sample post-annealed (irradiated and then annealed) at 600°C. All these observations are explained in terms of self-diffusion of Ag atoms on the Si(1 1 1) substrate, inter-diffusion of Ag and Si and phase formations in Ag and Si due to Ar ion irradiation.