G.E. Shatalov

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.

Calculated neutron transport verifications by integral 14 MeV-neutron source experiments with multiplying assemblies

This paper is devoted to a comparison of the experimental and calculated leakage neutron spectra and total neutron leakages from U, Th, Pb and Be spherical assemblies and also to a comparison of experimental and calculated reaction rates in U and Th assemblies. It was shown that the calculation with ENDF/B-IV data describes the shape of the neutron spectrum better for all considered materials. The full multiplication of neutrons for the U assembly is reproduced more accurately by the calculation with ENDL-75 data, and for Th, Be and Pb assemblies, by the calculation with ENDF/B-IV data.

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.

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.

Breeder and test blankets in ITER

Abstract An overview of ITER efforts is presented in this paper in the area of blanket development and ITER test program preparation. The design of ceramic and eutectic options of the driver blanket were developed aimed on a reliable operation and adequate tritium production. With an estimated net tritium breeding ratio of 0.8–0.95 ITER is able to achieve testing fluence goal of 1 MWa/m 2 during 8 years of technology phase operation. The test program was developed to ensure realistic extrapolation to DEMO on the basis of ITER experience. Submodules for the blanket, high heat flux components and material testing are to be inserted in especially designed test ports. Tests are provided for DEMO-relevant blanket design concepts.

DEMO blanket testing in ITER. Influence on reaching DEMO

ITER goal was specified as one step between now and the DEMO fusion reactor. One of the major issues is the tritium breeding blankets test relevant to future reactors. The major objectives of blanket modules (TBM) experiments in ITER are reduced in comparison with proposed test objectives in ITER-FDR. Thus, results of DEMO blanket designs testing in ITER will provide limited (but still useful) information that will need strong support from non-fusion facilities testing. The role of non-fusion tests is increased now to provide additional data required for DEMO blanket construction and qualification. A strategy of testing steps to DEMO blanket qualifications has to include parallel testing in ITER and in non-fusion devices. Experiments in fission reactors are able to provide essential data on materials radiation properties; tritium release, inventory and permeation; and thermomechanical behavior of the blanket breeder/multiplier. However, the volume in fission reactors is rather small and neutron spectra differ from the fusion reactor one. Nonetheless in the near future one depends primarily on fission reactor irradiation. The powerful accelerator based neutron source IFMIF could also provide useful information on radiation material properties. Plasma based neutron sources of different fusion devices could be the best choice for testing DEMO materials and blanket mock-ups. Timetable and costs of these devices are not clear now.

Attachment system for helium-cooled blanket of RF DEMO fusion reactor

The development of DEMO thermonuclear reactor is a part of Russian national program on the fusion process mastering. The DEMO-S (stationary thermonuclear reactor) should be the logic continuation of the ITER-type projects (pulse thermonuclear reactors) and the prototype for commercial power plants. DEMO reactor layout suggests to use the segmented blanket with mounting/dismounting procedure through the vacuum vessel vertical ports. Taking into account this layout the blanket attachment system has been developed and the present paper is devoted to this subject. The considered attachment system includes the lower and upper toroidal support assemblies which connect all the blanket segments in the enclosed ring. In its turn the lower support assemblies attached to the vacuum vessel through the system of hinged support pillars. The heights of support pillars for inboard and outboard blankets are selected so that to indemnify the blanket massif thermal expansions in vertical and radial directions. The support pillars have been calculated on strength taking into account the electromagnetic loads from the plasma disruptions and blanket mass. The selection of high-strength chromium steel as a structural material for the support pillars could be considered as the results of strength analysis. The conclusions on the possibility to apply this attachment system for fusion reactor blanket and the critical issues are contained in this paper too.

ITER shielding analysis

Abstract Radiation transport calculations have been carried out to aid in the design of the reference shielding and breeding blankets proposed for the International Thermonuclear Experimental Reactor (ITER). The results of analyses suggest that both blanket assemblies in combination with the surrounding vacuum vessel provide adequate shielding for the toroidal field coils and reduce the heating and damage to levels commensurate with design guidelines. The induced activation levels and decay heat generation in the breeding blanket—vacuum vessel may qualify the assemblies for disposal as Class C low level waste based on US regulations. The nuclear performance of the shielding around neutral beam injection ducts and in the vicinity of the divertor vacuum pumping ports was also investigated. Preliminary results suggest that the proposed neutral beam injection and divertor port shielding is satisfactory in both cases.

Properties and behaviour of 83Pb17Li eutectic

Abstract This article presents some test results of 83Pb17Li eutectic properties. This eutectic was produced by induction melting in the form of ingots with a volume of 30 l. It should be noted that inhomogeneity of eutectic composition and structure was revealed within one and the same ingot. Annealing at temperatures below melting point changes the micro structure i.e. increasing of the size of the β′-phase-particles takes place. The following values of eutectic thermal-physical properties were determined: enthalpy, specific heat capacity, thermal expansion, latent heat of melting. Some peculiarities of eutectic mechanical properties changing in the temperature range of 293 K to 473 K were investigated.

Materials issues in the design of the ITER first wall, blanket, and divertor

During the ITER conceptual design study, a property data base was assembled, the key issues were identified, and a comprehensive R&D plan was formulated to resolve these issues. The desired properties of candidate ITER divertor, first wall, and blanket materials are briefly reviewed, and the major materials issues are presented. Estimates of the influence of materials properties on the performance limits of the first wall, blanket, and divertor are presented.