N.N. Vasiliev

Russian DEMO-S reactor with continuous plasma burn

A conceptual study of DEMO-S steady state reactor is being carried out in RF. The main goals of the study are aimed at establishing physics and engineering basis and limitations of the reactor. It was assumed that DEMO is a next step after international thermonuclear experimental reactors (ITER) operation, and so conservative assumptions were used in the study based on ITER engineering. A steady state operation mode is recommended with fusion power of 2–3 GW and maximum first wall (FW) neutron load of 3–4 MW m−2. The reactor has to operate in advance plasma mode with high fraction of bootstrap current. The reversed shear regime and operation in H-mode are evaluated. A lifetime of 30–40 MW m−2 is assumed for the DEMO plant, with three to four changes of plasma facing components. The general reactor layout is determined. Two options of breeding blanket with ceramic and liquid lithium breeders are presented, supported by neutronic, thermalhydraulic and mechanical calculations. A conventional type of water or Li cooling divertor targets with maximum heat load of ∼10 MW m−2 was chosen. Heat to electricity conversion schemes are analysed for both helium and liquid metal coolant with net efficiency of 35–40%. Aspects of radioactive waste management are considered.

Russian DEMO plant study

Abstract A conceptual design study of the DEMO reactor is being carried out in Russia. The main efforts are concentrated on the study of steady state reactor DEMO-S. This reactor will be operated with an advanced plasma regime involving a high fraction of bootstrap current. A divertor concept with open liquid lithium as a plasma facing material was chosen. Two extreme approaches were analyzed: the first assumes that liquid lithium is loaded with direct heat flux from 100 to 150 MW m −2 ; the consumption of liquid lithium and a form of lithium surface are defined; in the second approach lithium plasma dynamics in the divertor is analysed using the 2-D MHD divertor code. With this approach heat loads are considerably decreased by the radiation. Some additional issues connected with utilization of vanadium alloys in the DEMO reactor, including environmental and safety aspects, material activation and refabrication were also analyzed.

Copper vapor laser application for surface monitoring of divertor and first wall in ITER

Abstract The availability of copper vapor laser system for surfaces surveying of International Thermonuclear Experimental Reactor (ITER) divertor and first wall during plasma discharge are justified. The construction concept of laser projector with image intensity amplification of the objects removed at a distance up to 20 m under intensive plasma background conditions is developed. The preliminary optical scheme for divertor and first wall surface monitoring from upper port is developed. Diffraction angle resolution at the object space is estimated (equal 1.18×10 −5 rad).The preliminary estimation of laser viewing system power parameters is carried out. Laser system operation under intensive plasma background conditions ( T th/ n =10 4 eV—typical thermonuclear temperature). Power relation signal/noise, calculated at the image intensity amplifier input is equal S / N =10 4 , i.e. plasma background radiation can be neglected in the considered scheme. Hence, the proposed technique application for monitoring of ITER discharge chamber internal surfaces is possible during running cycles, and in pauses especially. Experimental tests of laser system without scanning mirror and plasma noise background are carried out. Objects were located at a distance about 20 m from laser and amplifier. Probing laser had the next operating parameters: average radiation power, 10 W; pulse repetition rate, 10 kHz; output beam diameter, 14 mm. Linear resolution at objects plane about 1 mm was obtained. Degradation of resolution in comparison with estimated value—0.24 mm is related with insufficient degree of the image contrast, as the laser amplifier tube had unnormalized noise level. Method determined rate of the protecting cover erosion of surveyed surfaces is proposed. Under given measurement method application the laser viewing system can record decreasing of protection cover depth about 0.5 mm, and available value of protection cover erosion rate is equal 0.5 mm/ms. Objects surface observe methods are considered. Laser beam noninterlaced raster scanning on surveyed surface is proposed. The laser beam moving velocity is determined by required accuracy of the image reproduction. Pointing direction definition method based on checkpoint grids is offered. Quadratic pyramidal deepening center may be a checkpoint grid node of pointing direction system. Accuracy of pointing direction positioning at the object equals 2 mm provided an even decrease of protection cover thickness near the given deepening.

MAGRAS - facility for modelling of plasma facing beryllium sputtering and re-deposition

Abstract MAGRAS - magnetron facility for simulation of plasma facing beryllium sputtering and re-deposition is described. Optimization of its systems and components along with magnetron discharge phenomenology and its energy characteristics are considered. Studies perfonned at the facility can be used in designing plasma facing components of International Thermonuclear Experimental Reactor: in particular for their material selection, assessment of the life-time and determination of the near-wall plasma composition.

Possibility of fusion power reactor to transmute minor actinides of spent nuclear fuel

Abstract A possibility to use fusion power reactor (FPR) is considered for burning long-life elements of spent nuclear fuel in parallel with energy production. In this study a principal design of FPR blanket was examined for transmutation of long-life minor actinides (Np, Am, Cm). A production of minor actinide isotopes is equal to 20–30 kg/1 GW(e) year for now operating fission reactors, and their amounts will rise with the expected growth of fission reactor power. These isotopes have long-life time and can be dangerous in big amounts in future. Plutonium isotopes are not included in an assumption that they will be used in fission reactors. The major goals of the study were to determine FPR blanket composition corresponding to fast transmutation rate of actinides and tritium self-supply simultaneously. Tritium breeding ratio (TBR) was obtained at level 1.11 for water cooling and reached up 1.56 in variant with helium-cooled assemblies with Np nitride. It was concluded that rows with actinides from processed waste fuel should be arranged near the plasma first wall. Advantages of helium above water cooling are observed in the twice-increased loading of waste fissionable materials and essential increase of achievable TBR. Burnout of Np, Am, Cm would remain at a level ∼40–50% after 4 full power years.

Burning of Minor Actinides and Fission Products from Spent Nuclear Fuel of Power Plants in Dual-Purpose Fusion Reactor

The paper presents the results of analysis of transmutation of Minor Actinides (MA) and Fission Products (FP) from the Spent Nuclear Fuel (SNF) of nuclear power plants. The transmutation scenario includes repeating periods of neutron irradiation in dual-purpose Fusion Power Reactor-Tokamak (FPRT) with Deuterium-Tritium plasma as neutron source and periods of Partitioning and Reprocessing (P&R) of fuel between the irradiation cycles. FPRT is intended for power production and SNF burning. The investigations are based on the results of an RF DEMO-S project with a fusion power of 2.5u200aGW and a modified blanket including fuel assemblies with minor actinides (Np, Am, Cm) and selected fission products (I-129, Tc-99, Cs-135, Zr-93, Pd-107, Sn-126). All obtained results are in conceptual stage and detailed engineering investigations will be required in case of the concept approval for imbedding in the total nuclear fuel cycle.

Accumulation of deuterium in sputtered and re-deposited layers of Be and W under their simultaneous irradiation by deuterons

Abstract To understand the processes that take part under the simultaneous erosion of different plasma facing materials in a fusion reactor, experiments simulating ITER conditions were carried out, in which beryllium and tungsten targets were simultaneously exposed to deuteron fluxes causing their intensive erosion and re-deposition. This paper reports the results of a study of the retention of deuterium atoms in those layers, the composition and microstructure of re-deposited/sputtered beryllium and tungsten layers.

Modelling of hydrogen isotope ion interaction with beryllium plasma facing armour

To clarify the working capacity of beryllium plasma facing armour, beryllium targets were tested under bombardment with hydrogen and deuterium ions. The tests were performed under different experimental conditions, including those providing an intense re-deposition of erosion products. Chemical elements distribution, phase composition, microstructure of eroded and re-deposited Be-layers, and the accumulation of hydrogen isotopes in these layers are presented and discussed. Mutual re-deposition of Be and W erosion products and its effects on D-accumulation within the eroded areas and re-deposited layers are examined.

Fusion power plant for water desalination and reuse

Development of industry and agriculture demands a huge fresh water consumption. Exhaust of water sources together with pollution arises a difficult problem of population, industry, and agriculture water supply. Request for additional water supply in next 50 years is expected from industrial and agricultural sectors of many countries in the world. The presented study of fusion power plant for water desalination and reuse is aimed to widen a range of possible fusion industrial applications. Fusion offers a safe, long-term source of energy with abundant resources and major environmental advantages. Thus fusion can provide an attractive energy option to society in the next century. Fusion power tokamak reactor based on RF DEMO-S project [Proc. ISFNT-5 (2000) in press; Conceptual study of RF DEMO-S fusion reactor (2000)] was chosen as an energy source. A steady state operation mode is considered with thermal power of 4.0 GW. The reactor has to operate in steady-state plasma mode with high fraction of bootstrap current. Average plant availability of ∼0.7 is required. A conventional type of water cooled blanket is the first choice, helium or lithium coolants are under consideration. Desalination plant includes two units: reverse osmosis and distillation. Heat to electricity conversion schemes is optimized fresh water production and satisfy internal plant electricity demand The plant freshwater capacity is ∼6 000 000 m 3 per day. Fusion power plant of this capacity can provide a region of a million populations with fresh water, heat and electricity.

ITER shielding blanket

Abstract A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100–300 W cm−2. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used.