Neutronic performance comparation of MOX, nitride and metallic fuel based 25–100 MWe Pb–Bi cooled long life fast reactors without on-site refuelling

In this paper the performance of 25–100 MWe Pb–Bi cooled long life fast reactors based on three type of fuels: MOX, Nitride and Metal are compared and discussed. In general MOX fuel (UO₂–PuO₂) has lower atomic density compared to the nitride or metal fuel, but MOX fuel has some advantages such as higher Doppler coefficient, high melting point and availability. Nitride fuel has advantages such as high density, high thermal conductivity, and high melting point, but need N-15 to avoid C-14 problems.

The results show that nitride fuel as well as MOX fuel can be used to develop 25–100 MWe (75–300 MWth) Pb–Bi cooled long life reactors without on-site fuelling. The results show that nitride fuels have more superior neutronic characteristics compared to that of MOX fuel due to higher density. However, in the large power level both fuels can be easily applied. In lower power level the MOX fuel need higher fuel volume fraction to reach the comparable target of nitride fuelled cores.     

Phase relations and thermodynamic properties of the uranium-nitrogen system

The present state of knowledge of the phase relations and thermodynamic properties of the uranium-nitrogen system is given. Emphasis is placed on the nonstoichiometric and thermodynamic properties of uranium sesquinitride. The present paper consists of two parts. The first part is on the phase relations; an equilibrium phase diagram is proposed, and the relationship between structure and lattice parameter for the α-U₂N₃-UN₂ system is shown. The second part deals with thermodynamic properties of UN and α-U₂N₃; the data on decomposition pressures, specific heats, and heats and free energies of formation are summarized and evaluated.     

Safety performance comparation of MOX, nitride and metallic fuel based 25–100 MWe Pb–Bi cooled long life fast reactors without on-site refuelling

In this paper the safety performance of 25–100 MWe Pb–Bi cooled long life fast reactors based on three types of fuels: MOX, nitride and metal is compared and discussed. In the fourth generation NPP paradigm, especially for Pb–Bi cooled fast reactors, inherent safety capability is necessary against some standard accidents such as unprotected loss of flow (ULOF), unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS). Selection of fuel type will have important impact on the overall system safety performance.

The results of safety analysis of long life Pb–Bi cooled fast reactors without on-site fuelling using nitride, MOX and metal fuel have been performed. The reactors show the inherent safety pattern with enough safety margins during ULOF and UTOP accidents. For MOX fuelled reactors, ULOF accident is more severe than UTOP accident while for nitride fuelled cores UTOP accident may push power much higher than that comparable MOX fuelled cores.     

Zr-rich layers electrodeposited onto stainless steel cladding during the electrorefining of EBR-II fuel

Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U–Zr alloy fuel elements irradiated in the Experimental Breeder Reactor II (EBR-II). We report the first metallographic characterization of cladding hull remains for the electrometallurgical treatment of spent metallic fuel. During the electrorefining process, Zr-rich layers, with some U, deposit on all exposed surfaces of irradiated cladding segments (hulls) that originally contained the fuel alloy that was being treated. In some cases, not only was residual Zr (and U) found inside the cladding hulls, but a Zr-rind was also observed near the interior cladding hull surface. The Zr-rind was originally formed during the fuel casting process on the fuel slug. The observation of Zr deposits on all exposed cladding surfaces is explained with thermodynamic principles, when two conditions are met. These conditions are partial oxidation of Zr and the presence of residual uranium in the hulls when the electrorefining experiment is terminated. Comparisons are made between the structure of the initial irradiated fuel before electrorefining and the morphology of the material remaining in the cladding hulls after electrorefining.     

Thermodynamic evaluation of the Nb–O–Zr system

Thermodynamic properties of the ternary system Nb–O–Zr have been evaluated by means of the CALPHAD (CALculation of PHAse Diagrams) method. The pertinent experimental data are surveyed and the thermodynamic models based on the previous assessments of the binary systems Nb–O, Nb–Zr and O–Zr are delineated. The results of our computations indicate that the models describe the zirconium rich portion of the ternary phase diagram satisfactorily, however, in the niobium rich part, the calculations differ from the experimental data and should be verified by new experiments.     

Thermochemical Stability, Radiation Testing, Fabrication, and Reprocessing of Mononitride Fuel

A thermodynamic analysis and experimental investigations have shown that mononitride fuel is thermochemically stable up to 1973–2073 K, at which temperature the equilibrium vapor pressure of nitrogen does not exceed 4.5·10^–7–2.1·10^–6 MPa. It is concluded on the basis of a generalization of the data from radiation testing of mononitride fuel with burnup up to 9–10% in fast and 16.8% in thermal reactors with lineal power density from 400 to 1300 W/cm that it should operate reliably in fuel elements with helium and liquid-metal sublayers. The requirement for the impurity (oxygen and carbon) content in it is formulated. When both oxygen and carbon impurities are present simultaneously in mononitride, the mass fraction of each should not exceed 0.15%. The methods for fabricating mononitride fuel are determined by the final product of the reprocessing of irradiated fuel. Consequently, methods for fabricating mixed nitride fuel from oxides and metals are now being developed.     

Uranium nitride fuels in superheated steam

Uranium mononitride (UN) pellets of different densities were subjected to a superheated steam/argon mixture at atmospheric pressure to evaluate their resistance to hydrolysis. Complete degradation of pure UN pellets was obtained within 1 h in 0.50 bar steam at 500 °C. The identified reaction products were uranium dioxide, ammonia, and hydrogen gas, with no detectable amounts of nitrogen oxides formed. However, the reaction could not be carried to completion, and the presence of uranium sesquinitride and higher uranium oxides or uranium oxynitrides in the solid residue is indicated. Evolution of elemental nitrogen was seen in connection with very high reaction rates. The porosity of the pellets was identified as the most important factor determining reaction rates at 400–425 °C, and it is suggested that in dense pellets, cracking due to internal volume increase initiates a transition from slow surface corrosion to pellet disintegration. The implications for the use of nitride fuels in light water reactors (LWR) are discussed, with some observations concerning hydrolysis as a method for ¹⁵N recovery from isotopically enriched spent nitride fuel.     

Thermodynamic modeling of the Mg-Pu and Cu-Pu systems

The thermodynamic assessment of the Mg-Pu and Cu-Pu systems was carried out by using the calculation of phase diagrams (CALPHAD) method on the basis of experimental data including thermodynamic properties and phase equilibria. The Gibbs free energies of the liquid, fcc, hcp, αPu, βPu, γPu, δPu, δ′Pu, and εPu phases were described by the subregular solution model with a Redlich-Kister equation, and those of the intermetallic compounds in the Mg-Pu and Cu-Pu binary systems were described by the sublattice model. A consistent set of thermodynamic parameters were derived for describing the Gibbs free energies of solution phases and intermetallic compounds in the Mg-Pu and Cu-Pu binary systems. An agreement between the calculated results and experimental data is obtained.     

Measurement of the specific heat of Zr–40 wt%U metallic fuel

The specific heat capacities of un-irradiated and irradiated metallic Zr–40 wt%U fuel have been measured between 50 °C and 1000 °C with a differential scanning calorimetry. The irradiated fuels have three different burnup levels of 0.38, 0.70 and 0.92 g-fission product (FP)/cm³. The measured specific heat for the un-irradiated fuel is representative and consistent with the values estimated from the Neumann–Kopp rule. The irradiated fuels exhibited a complicated behavior of the heat capacities. The unique characteristics of the specific heat capacities can be explained by the recovery of radiation damage, the formation of fission gas bubbles and fission gas release, and a phase transition in the irradiated fuels. An examination of the microstructure revealed that multiple large bubbles were formed in the irradiated fuel during specific heat measurement. The measured specific heat is expected to enable us to estimate the stored energy in the metallic fuel during certain accident scenarios and to determine the thermal conductivity of zirconium–uranium metallic fuel.     

Phase equilibrium of PuO2−x − Pu2O3 based on first-principles calculations and configurational entropy change

Combination of an oxygen vacancy formation energy calculated using first-principles approach and the configurational entropy change treated within the framework of statistical mechanics gives an expression of the Gibbs free energy at large deviation from stoichiometry of plutonium oxide PuO₂. An oxygen vacancy formation energy 4.20 eV derived from our previously first-principles calculation was used to evaluate the Gibbs free energy change due to oxygen vacancies in the crystal. The oxygen partial pressures then can be evaluated from the change of the free energy with two fitting parameters (a vacancy-vacancy interaction energy and vibration entropy change due to induced vacancies). Derived thermodynamic expression for the free energy based on the SGTE thermodynamic data for the stoichiometric PuO₂ and the Pu₂O₃ compounds was further incorporated into the CALPHAD modeling, then phase equilibrium between the stoichiometric Pu₂O₃ and non-stoichiometric PuO_2−x were reproduced.     

Pyrochemical reduction of uranium dioxide and plutonium dioxide by lithium metal

The lithium reduction process has been developed to apply a pyrochemical recycle process for oxide fuels. This process uses lithium metal as a reductant to convert oxides of actinide elements to metal. Lithium oxide generated in the reduction would be dissolved in a molten lithium chloride bath to enhance reduction. In this work, the solubility of Li₂O in LiCl was measured to be 8.8 wt% at 650 °C. Uranium dioxide was reduced by Li with no intermediate products and formed porous metal. Plutonium dioxide including 3% of americium dioxide was also reduced and formed molten metal. Reduction of PuO₂ to metal also occurred even when the concentration of lithium oxide was just under saturation. This result indicates that the reduction proceeds more easily than the prediction based on the Gibbs free energy of formation. Americium dioxide was also reduced at 1.8 wt% lithium oxide, but was hardly reduced at 8.8 wt%.     

Coupled neutronics/thermal hydraulics evaluation for thorium based fuels in thermal spectrum SCWR

Thorium can supplement the current limited reserves of uranium. In current study, analyses are performed for thorium based fuels in thermal neutron spectrum Super Critical Water Reactor (SCWR). Thorium based fuels are studied in two roles. First role being replacement of conventional uranium dioxide fuel while the other being burner of Reactor Grade Plutonium (RG-Pu) in thermal neutron spectrum SCWR. Coupled neutron physics/thermal hydraulics analyses are performed due to large density variation of coolant over the active fuel length. Analyses reveal that thorium-uranium MOX fuels lead to smaller burnup values as compared to equivalent enriched uranium dioxide but possess the advantage of smaller excess reactivity at Beginning of Life (BOL). This can lead to savings in the form of Burnable Poisons (BP). Smaller fuel average temperature values are obtained for thorium-uranium MOX fuels as compared to uranium dioxide fuel option. Coated fuel option utilizing mixed thorium-uranium mono nitride fuel can help further decrease fuel average temperature values for thorium based fuels. U-233, produced in thorium uranium fuels, contribution towards fission energy produced is smaller as compared to plutonium produced in conventional uranium dioxide fuel. In terms of proliferation resistance, approximately 40% less quantity of plutonium is produced for thorium-uranium MOX fuels (for studied compositions) as compared to equivalent enriched uranium dioxide fuel. But, there is not much difference between the discharged plutonium vector compositions. Thorium–Plutonium based fuels lead to significantly harder spectrum which results in larger spread in radial power density and eventually causes larger values for thermal hydraulic parameters like fuel and clad temperature. Due to almost no production of plutonium, thorium based fuels can be a very good option to burn RG-Pu in thermal spectrum SCWR. Thorium based fuels destroyed almost 74% initially loaded RG-Pu as compared to 60% for uranium based MOX. HEU based thorium fuels can be a very good option for replacing conventional uranium dioxide fuels as very small quantities of plutonium is produced. This option, although, has regulatory issues due to use of HEU material.     

Thermodynamics of high-temperature nuclear fuel

A method for performing a thermodynamic analysis of the high-temperature nuclear fuel using the ASTA computer program is substantiated. Calculations of the chemical composition and pressure of the gas phase of the ternary systems U-O-C and Pu-O-C are performed. The results obtained are compared with existing experimental data and theoretical studies performed by other authors. The results show that the entropy factor apparently plays an appreciable role in the thermodynamics of the systems studied. A comparative analysis of micropellets with uranium and plutonium fuel is performed. An estimate of the diffusion kinetics of the chemical interaction in micropellets is given as substantiation of the application of the methods of equilibrium thermodynamics for calculating the chemical and phase composition of the nuclear fuel.__________Translated from Atomnaya Énergiya, Vol. 98, No. 1, pp. 36–44, January, 2005.     

Fission gas production in reactor fuels including the effects of ternary fission

An understanding of gas bubble formation and migration in nuclear fuel and its impacts on fuel and cladding materials requires knowledge of the isotopic composition of the gases and their generation rates. In this paper, we present results of simulations for the production of the dominant noble gases (helium, xenon, krypton) in nuclear fuels for different reactor core configurations and fuel compositions. The calculations were performed using detailed nuclear burn simulations with Monte Carlo nuclear transport, and included ternary fission to ensure an accurate treatment of helium production. For all reactor designs and fuels considered xenon was found to be the most dominant gas produced. Variation in the composition of fission gases is quantified for: (1) the burn time, (2) the composition of the fuel, and (3) the neutron energy spectrum. These three factors determine the relative fraction of each gas and its transmutation into or from stable gas by subsequent neutron capture.     

快堆中不同核燃料类型的长寿期性能评价

建立了典型的快堆六角形栅元堆芯模型,研究了多种类型的燃料在快中子能谱辐照环境下经过较长时间辐照后的性能,对不同燃料堆芯在运行寿期末的乏燃料组成成分进行了分析.结果表明,在栅元结构完全一样且初始剩余反应性基本相同的情况下,燃料反应性损失从小到大的顺序是:金属燃料＜氮化物燃料＜碳化物燃料<氧化物燃料;在整个寿期中,使用Pu驱动的燃料比使用235U驱动的燃料反应性下降得慢;金属燃料寿期末乏燃料中按初始装载燃料质量平均后的超铀核素的质量最小,其他依次为氧化物<氮化物<碳化物;由于初始装载量的增多,使用Pu驱动的燃料寿期末乏燃料超铀核素的总量比使用235U驱动的燃料多,同时,乏燃料Pu中的易裂变同位素的份额比235U驱动燃料的少.从中子学角度考虑,UZr燃料是比较理想的长寿命快堆堆芯燃料类型.     

Analysis of high burnup fuel behavior in Halden reactor by FEMAXI–V code

The author developed a code FEMAXI–V to analyze the behaviors of high burnup LWR fuels. FEMAXI–V succeeded the basic structure of code FEMAXI–IV, and incorporated such new models and functions as fuel thermal conductivity degradation with burnup, alliance with burnup analysis code which gives radial power profile and fast neutron flux, etc. In the present analysis, coolant conditions, detailed power histories and specifications of the fuel rods DH and DK of IFA-519.9 irradiated in Halden reactor were input, and calculated rod internal pressures were compared with experimental data for the range of 25–93 MWd kg⁻¹ UO₂, and factors affecting pellet temperature were discussed. Also some sensitivity studies were conducted with respect to the effect of swelling rate and grain growth. As a result, it is found that the prediction is sensitive to the models of thermal conductivity and swelling rate of fuel, and FEMAXI–V analytical system proved to give a reasonable prediction even in the high burnup region.     

Synergistic hydrogen production by nuclear-heated steam reforming of fossil fuels

Processes and technologies to produce hydrogen synergistically by the nuclear-heated steam reforming reaction of fossil fuels are reviewed. Formulas of chemical reactions, required heats for reactions, saving of fuel consumption, reduction of carbon dioxide emission, and possible processes are investigated for such fossil fuels as natural gas, petroleum and coal.

In this investigation, examined are the steam reforming processes using the “membrane reformer” and adopting the recirculation of reaction products in a closed loop configuration. The recirculation-type membrane reformer process is considered to be the most advantageous among various synergistic hydrogen production processes. Typical merits of this process are; nuclear heat supply at medium temperature around 550°C, compact plant size and membrane area for hydrogen production, efficient conversion of a feed fossil fuel, appreciable reduction of carbon dioxide emission, high purity hydrogen without any additional process, and ease of separating carbon dioxide for future sequestration requirements.

The synergistic hydrogen production using fossil fuels and nuclear energy can be an effective solution in this century for the world which has to use fossil fuels to some extent, according to various estimates of global energy supply, while reducing carbon dioxide emission.     

Calculations of thermodynamic properties of PuO2 by the first-principles and lattice vibration

Plutonium dioxide (PuO₂) is a key compound of mixed oxide fuel (MOX fuel). To predict the thermal properties of PuO₂ at high temperature, it is important to understand the properties of MOX fuel. In this study, thermodynamic properties of PuO₂ were evaluated by coupling of first-principles and lattice dynamics calculation. Cohesive energy was estimated from first-principles calculations, and the contribution of lattice vibration to total energy was evaluated by phonon calculations. Thermodynamic properties such as volume thermal expansion, bulk modulus and specific heat of PuO₂ were investigated up to 1500 K.     

Evaluation on chemical state of irradiated rock-like oxide fuels by nuclear and chemical modeling

Chemical forms of fission products in irradiated ROX fuels were calculated by the SOLGASMX-PV code, and the resultant phase equilibrium and the oxygen potential in the fuel were evaluated in order to assess the irradiation behavior of the ROX fuels. For the ROX fuel with reactor grade Pu, the oxygen potential increased to about −140 kJ mol⁻¹ at EOL when all the Pu in the fresh fuel was tetravalent. In the case of fresh fuel which was partially reduced with the [Pu⁺³]/[Pu⁺⁴]=10/90, the oxygen potential increase was suppressed to about −400 kJ mol⁻¹. On the other hand, the oxygen potential of the ROX fuel with weapon grade Pu never exceeded the value of about −400 kJ mol⁻¹. The difference of oxygen potentials was caused by difference of Am amount produced by Pu conversion. The oxygen potential of the irradiated fuel was controlled by the phase equilibria among FPs. The equilibrium between metallic Mo and MoO₂ controlled the oxygen potential to about −400 kJ mol⁻¹.     

Thermodynamic Properties of U0.718Ce0.282O2±x

An experimental study of the structure and thermodynamic properties of a sample of U_0.718Ce_0.282O_2±x , which is an analogue of uranium–plutonium oxide, is performed to study the effect of adding cerium oxide on the properties of oxide fuel. The experimental specimens were obtained by powder metallurgy. x-Ray diffraction analysis of sintered tablets showed the formation of an equilibrium one-phase solid substitution solution based on UO₂ with substoichiometric composition. The potentiometric method of the emf of a solid-electrolyte galvanic cell in the range 850–1050°C and solid-phase coulometric titration with working temperature 1000°C were used to study the dependence of the oxygen potential of uranium-cerium oxide on the oxygen/metal ratio and the temperature. The concentration dependences of the enthalpy and the entropy of dissolution of oxygen in uranium-cerium oxide are calculated and constructed and it is determined that these dependences vary strongly at near-stoichiometric composition. The results can be used to analyze the properties of uranium–plutonium oxide fuel.