The experimental results obtained within a joint international research e¨ort regarding the formation of condensed combustion products from nanoaluminum-based solid propellants (SPs) are reported. Data on the size, structure, chemical composition, and quantity of condensed combustion products (CCPs) as well as conditions of their formation are discussed. On the basis of the collected experimental data, a general physical picture of condensed combustion products formation is portrayed. The results of this study allow carrying out the analysis of good quality propellants using nanoaluminum.

A scheme for circulating coolant and cooling the core that has advantages over the designs of similar nuclear power systems is proposed for light-water reactors with supercritical coolant parameters and a fast-resonance neutron spectrum. A negative void coefficient of reactivity is obtained for the entire run of a fuel assembly without building a blanket. A more uniform distribution of the energy release over the core volume is achieved without using complicated fuel-enrichment schemes. The nonuniformity of the coolant temperature distribution at the core exit is decreased. The fuel assemblies operate with a much lower temperature drop over the core height. The core has a small reactivity excess on burnup and a BR of about 1, for which the most difficult operating regimes (flooding with cold water) can be handled with standard means (placement of absorbing organs of the safety and control system in ~2/3 of the fuel assemblies).The cost of the equipment of the steam-generating facility comprises 70% of the capital costs of constructing the power-generating units of a nuclear power plant. If the thermal scheme is complicated and the coolant parameters are low, the efficiency of nuclear power plants with VVÉR and RBMK reactors is 33-35% as compared with 50% for thermoelectric power plants. The main problems of new designs are to decrease the specific capital investments in a facility, use simplified schemes and processes, and decrease the construction time while ensuring a high level of safety. Reactors with light-water coolant at supercritical pressure permit solving these problems most completely. A simple thermal scheme (steam from the reactor flows directly onto the turbine) and the elimination of a large amount of expensive equipment (steam generators, pumps, pipelines, second-loop fixtures) decrease the metal content by 60%. High steam parameters (pressure ~25 MPa, temperature 535-545°C) and a single-loop scheme will make it possible to obtain a facility efficiency of 44%. Reducing the required amount of coolant in the core will make it possible to arrange the fuel elements in closely-spaced lattices, and the reactor will have a fast neutron spectrum with a breeding ratio of about 1.The GIF international program for developing generation-IV reactors has been ongoing since 2000. In this program, water-cooled reactors with supercritical-pressure coolant will also be among the advanced future reactors.Japanese scientists are working most intensively in this field. They have proposed reactor designs with a thermal neutron spectrum SCPR [1, 2] and a fast neutron spectrum SCFR [3]. The latter reactor will use a mixed fuel based on depleted uranium oxide and uranium oxide enriched with weapons plutonium. The core consists of 270 fuel assemblies with uranium and plutonium dioxide fuel and 163 fuel assemblies with fuel consisting of depleted uranium (blanket). The reactor vessel and blanket are cooled with 280°C and 25 MPa water moving from top to bottom. At the bottom, the flows are mixed and cool the core in the up c...

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Computational results obtained for fuel assemblies and a core with uni-and bidirectional coolant flow are presented for a water-cooled reactor with a thermal neutron spectrum. It is shown that a bidirectional scheme for cooling fuel assemblies has advantages over a unidirectional scheme and holds promise for Gen IV water-cooled reactors with supercritical coolant pressure, which make it possible to perfect the technology for closing and drawing thorium into the fuel cycle.The Gen IV water-cooled reactors with supercritical-pressure coolant are considered to be promising and should come on line in approximately 2030. Reactors with a thermal neutron spectrum are being developed abroad as well as in our country as a solution to the near-term problem of replacing light-water reactors and will be followed later on by reactors with a closely spaced fuel-element lattice and a fast neutron spectrum [1,2]. Unidirectional cooling schemes where the coolant is heated as it moves in the core from bottom to top have been adopted for such designs. Since the heating amounts to 230°C in thermal and 250°C in fast reactors, even small variations of the energy-release distribution of the fuel elements result in large differences of the coolant temperature at the exit and the cladding temperature of the fuel elements. In a reactor with a thermal neutron spectrum, the cold moderator is confined inside tubes -"water elements" between which the fuel elements, cooled by the ascending coolant, are arranged in a closely spaced lattice. A difficulty is smoothing the energy-release field, since the moderator is substantially heated and its density varies along the height of the core.For reactors with a fast neutron spectrum, it is suggested that a bidirectional cooling scheme be used to smooth the energy release over the core height [3,4].In the present article, we present the results of a physical calculation, performed using different program systems, and a comparative analysis of a reactor with a thermal neutron spectrum with uni-and bidirectional cooling schemes. Uranium-plutonium-thorium fuel loads are considered.Fuel Assemblies with Cylindrical Water Elements (unidirectional cooling scheme). It has been suggested that a cylindrical channel with two coaxially arranged shells separated by a gap filled with water ( Fig. 1) be used to decrease the heating of the moderator over the core height [5]. Zirconium can be used for the inner tube of the water elements and corrosion-resistance steel for the outer tube. A variant where the outer tube consists of zirconium up to 180 cm from the bottom and steel above this level is also under consideration.The working model consists of a full-scale description of the reactor in the radial and vertical directions taking the end and radial reflectors into account. The VVER-1500 core was used as a prototype for the core, the main difference being the design of the fuel assemblies and the power. The basic characteristics of the core and fuel-assembly variants are as follows:

RUTA-70 model for simulations with the internally coupled codes DYN3D/ATHLET and DYN3D/RELAP5 was developed. A 3-D power distribution in the core is calculated by DYN3D with thermal-hydraulic feedback from the system codes. A steady-state corresponding to the full reactor power and an accident scenario initiated by failure of all primary coolant pumps were simulated with the DYN3D/ATHLET and DYN3D/RELAP5 coupled code systems to verify these codes. The compared coupled codes give close predictions for the initial and final states of the simulated accident but not for the transition between them. The observed deviations are explained by differences in the subcooled boiling models of the employed versions of ATHLET and RELAP5. Nevertheless, both simulations confirm a high level of the reactor inherent safety. The allowed safety margins were not reached.