Development of fatigue life criteria for experimental fusion reactor first-wall structures

An approach to the rational design of fusion reactor first-wall structures against fatigue crack growth is proposed. The approach is motivated by microstructural observations of fatigue crack growth enhancement in unirradiated materials due to volumetric damage ahead of a propagating crack. Examples are cited that illustrate the effect of mean stress on void nucleation and coalescence, which represent the dominant form of volumetric damage at low temperature, and of grain boundary sliding and creep cavitation, which are the dominant volumetric damage mechanisms at high temperature. The analogy is then drawn between these forms of fatigue crack growth enhancement and those promoted by irradiation exposure in the fusion reactor environment, such as helium embrittlement and atomic displacement. An enhanced strain range is suggested as a macroscopic measure of the reduction in fatigue life due to the higher fatigue crack growth rates. The enhanced strain range permits a separation of volumetric and cyclic effects, and assists in the assignment of rational design factors to each effect. A series of experiments are outlined which should provide the numerical values of the parameters for the enhanced strain range.     

Physical sputtering model for fusion reactor first-wall materials

A model has been developed to provide energy-dependent physical sputter yields for various plasma particles (ions and neutrals) incident on candidate first-wall materials. The physical sputter yield is expressed in terms of the atomic and mass numbers of the projectile and target atoms, the surface binding energy of the targets and the energy of the incident particle. The general shapes of the yield curves are based on theoretical models, whereas the magnitudes of the yields are derived primarily from experimental data. The model applies to both high- and low-Z incident particles bombarding high- and low-Z wall materials. Although the model was developed for metal first-wall materials, it has been extended, with minor modifications, to predict physical sputter yields of several stable compound wall materials. A comparison of predicted yields with available experimental data has been made for a number of candidate first-wall materials.     

Fusion reactor first-wall cooling for very high energy fluxes

A study has been made of the feasibility of a protective first-wall shield between the plasma and the containment vessel for early experimental controlled thermonuclear fusion machines. The proposed first-wall shield is a water-cooled array of thin-walled tubes designed to take very high local energy fluxes originating from the neutral beam injectors. Detailed computer calculations reveal that heat flux capabilities of 3300 W/cm² are possible with first-wall shield sections made up of tubes 1 m long of Ta-10W alloy (with tubes of 10 mm i.d. and tube wall thickness of 0.5 mm) with a structural safety factor of about four. Required pumping powers on the order of 1 MW/m² of first-wall area exposed to these high energy fluxes are predicted for flow in the non-boiling regime. If operation in the subcooled nucleate boiling regime can be achieved without oscillations or instabilities, the required pumping power is shown to decrease by about an order of magnitude.     

A novel digital neutron flux monitor for international thermonuclear experimental reactor

A novel full-digital real-time neutron flux monitor(NFM) has been developed for the International Thermonuclear Experimental Reactor.A measurement range of 10~9 counts per second is achieved with 3 different sensitive fission chambers.The Counting mode and Campbelling mode have been combined as a means to achieve higher measurement range.The system is based on high speed as well as parallel and pipeline processing of the field programmable gate array and has the ability to upload raw-data of analog-to-digital converter in real-time through the PXIe platform.With the advantages of the measurement range,real time performance and the ability of raw-data uploading,the digital NFM has been tested in HL-2 A experiments and reflected good experimental performance.     

Heat transfer in inertial confinement fusion reactor systems

The short time and deposition distance for the energy from inertial fusion products results in local peak power densities on the order of 10¹⁸ W/m³. This paper presents an overview of the various inertial fusion reactor designs which attempt to reduce these peak power intensities and describes the heat transfer considerations for each design.     

Thermal-hydraulic design of water-cooled pressure tube blanket for a fusion driven subcritical reactor

A Water-cooled Pressure Tube Energy production blanket (WPTE) for fusion driven subcritical reactor has been designed to achieve 3000 MW thermal power with self-sustaining tritium cycle. Pressurized water has great advantages in energy production; however the high pressure may cause some severe structural design issues. This paper proposes a new concept of water-cooled blanket. To solve the problem of the high pressure of the coolant, the pressure tube was adopted in the design and in the meantime, the thickness of the first wall can be significantly reduced as result of adopting pressure tube. The numerically simulating and calculating of temperature, stress distribution and flow analyses were carried out and the feasibility of using water as coolant was discussed. The results demonstrated the engineering feasibility of the water-cooled fusion–fission hybrid reactor blanket module.     

Selection of a toroidal fusion reactor concept for a magnetic fusion production reactor

The basic fusion driver requirements of a toroidal materials production reactor are considered. The tokamak, stellarator, bumpy torus, and reversed-field pinch are compared with regard to their demonstrated performance, probable near-term development, and potential advantages and disadvantages if used as reactors for materials production. Of the candidate fusion drivers, the tokamak is determined to be the most viable for a near-term production reactor. Four tokamak reactor concepts (TORFA/FED-R, AFTR/ZEPHYR, Riggatron, and Superconducting Coil) of approximately 500-MW fusion power are compared with regard to their demands on plasma performance, required fusion technology development, and blanket configuration characteristics. Because of its relatively moderate requirements on fusion plasma physics and technology development, as well as its superior configuration of production blankets, the TORFA/FED-R type of reactor operating with a fusion power gain of about 3 is found to be the most suitable tokamak candidate for implementation as a near-term production reactor.This paper represents work carried out from 1980 to 1982 and was in draft form in 1982. It was received for publication with only minor editing from its 1982 version (except for Tables II and III and Fig. 1), explaining the fact that some of the material is dated.     

The neutronics studies of a fusion fission hybrid reactor using pressure tube blankets

In this paper, a fusion fission hybrid reactor used for energy producing is proposed based on the situation of nuclear power in China. The pressurized light water is applied as the coolant. The fuel assemblies are loaded in the pressure tubes with a modular type structure. The neutronics analysis is performed to get the suitable design and prove the feasibility. The energy multiplication and tritium self-sustaining are evaluated. The neutron load is also cared. From different candidates, the PWR spent fuel is selected as the feed fuel. The results show that the hybrid reactor can meet the expected reactor core lifetime of 5 years with 1000 MWe power output. Two ways are discussed including burning the discharged PWR spent fuel and burning the reprocessed plutonium. The energy multiplication is big enough and the tritium can be self-sustaining for both of the two ways. The neutron wall load in the operating time is kept smaller than the one of ITER. The way to use the reprocessed plutonium brings low neutron wall load, but also brings additional difficulties in operating the hybrid reactor. The way to use the discharged spent fuel is proposed to be a better choice currently.     

Transport processes in an inertial confinement fusion reactor

The quasi one-dimensional method previously developed for calculating a transient, compressible, viscous flow through a complex array of tubes or jets was extended to include heat and mass exchange between the fluid and the jets. The application was again the impulsive crossflow of a lithium plasma through a close-packed annular arrangement of liquid jets, a problem that arises in the deisgn of inertial confinement fusion reactors. It was found that the peak hoop stress in the first wall of the reactor may derive from the direct impact of the plasma, rather than from the subsequent impact of the jets or fragments thereof. Depending on conditions in the cavity, the peak wall stress was calculated to be up to 4 times less when heat and mass transfer were accounted for. The sensitivity of the design to key parameters was established.     

Determination of the neutron flux density at an unknown reactor power

In the present paper, relationships are derived which enable, by using an additional measurement, to calibrate the neutron flux density measured at an unknown low reactor power to the power of 1 MW. The measurements can be carried out with activation detectors using well-known methods, see e.g. (ASTM, 1977, 1980; Zijp, 1984). Constant neutron flux density at a power of 1 MW is assumed during the measurements and a so-called “one-point reactor” model is employed.     

Structural considerations for a Tokamak fusion reactor

Several preliminary structural analyses are presented which validate a design for the experimental power reactor. Three components are singled out as requiring special attention: the magnetic coils, the blanket support structure, and the blanket modules. Repeated loading of a coil structure by magnetic forces should produce only linear elastic deformation. An analysis for minimum preload necessary to ensure this is presented. Using axisymmetric thin shell theory, a stress analysis of the blanket support structure is described. To account for the welded ring structure, a perforated plate analysis is used to compute the structural displacements and the ligament stresses. Temperature distributions and thermal stresses in the blanket module are determined using both finite element and analytical analysis. The stresses are all acceptable, including the effects produced by creep and fatigue. Thermal stress in the liner produced by a nonlinear temperature gradient is also shown to be acceptable.     

Helium cooling circuits for a fusion reactor

A thermal and hydraulic analysis of the complete cooling circuit of a conceptual helium-cooled fusion reactor is presented. A manifolding analysis is first applied to rows of blanket cells. It is shown that duct diameters must be of the order of 0.15 m or greater; if they are too small not only is the pressure drop too great but some blanket cells are overcooled at the expense of others which overheat.The complete system consisting of many rows of cells and the interconnecting ductwork, together with pumps and heat exchangers, is then analysed. Design option regions are drawn for a variety of operating parameters. These regions are determined by maximum pumping power fraction, minimum economic wall loading, minimum thermodynamic efficiency and maximum temperature. Because the maximum operating temperature depends on the material of construction, different regions can be drawn for different materials. The size of the region, and with it the maximum wall loading, decreases as the pressure and the coolant ducting diameter decrease.     

Neutron multiplier alternatives for fusion reactor blankets

A proposal is made to replace the neutron multiplier in fusion reactor blankets by an efficient moderator (⁷LiH or ⁷LiD). The advantageous effect of the intensified neutron-energy degradation is due to the $\text{1}\text{v}$ character of the main tritium-producing reaction. The slowing-down medium is designed to be the source of moderated neutrons for the surrounding Li region where the most of the tritium is to be produced. The surplus tritium produced remains stored in the moderator zone. Some preliminary calculations illustrating the above concept were carried out, and the neutron flux and tritium production distributions are presented. Indications regarding further studies are also suggested.     

Liquid lithium divertor system for fusion reactor

One of the most critical issues for the steady state fusion reactor is the heat flux in the divertor target. This paper proposes a liquid lithium divertor system to solve this problem. The proposed divertor system consists of a liquid lithium target, an evaporation chamber and a differential evacuation chamber. The heat coming from the fusion plasma along the divertor leg is removed by evaporation of lithium. The lithium vapor is condensed on the wall and is circulated with a pump. The coolant temperature for the wall is high enough to drive a power generator. Narrow slits along the divertor leg and the differential evacuation chamber reduce leakage of lithium vapor to the plasma chamber. A preliminary estimation predicts that the lithium ion density in the core plasma is lower than the plasma density.     

Approach for fusion reactor safety and fusion safety work in Russia

Main directions of work on experimental fusion reactors safety assurance in Russia are given. Work on safety includes: the elaboration of the main criteria and principles of safety assurance, the development of the first priority standards in safety on the basis of the fission experience and international safety documents requirements, fusion reactor safety analysis, and work to provide a base for the standards development and for the safety analysis activity. The results of some work on fusion safety are presented. They include: assessments of safety and reliability of Liquid Metal Cooling System draft design, evaluations of the buildings and equipment response on external dynamic influences, and analysis of radiological situation in th environment as a result of tritium-containing dust release.     

聚变反应堆含氚废水处理系统的设计

以联合电解催化交换-气相色谱(CECE-GC)为技术路线基础,对聚变反应堆(International thermonuclearexperimental reactor,ITER)含氚废水处理系统(Water detritiation system,WDS)进行了总体设计和主要子系统的设计.与目前的重水提氚演示系统相比,ITER-WDS的不同之处在于不使用氢氧复合器,不采用碱性电解池而使用固体聚合物电解池(Solid polymer electrode,SPE),增加了Pd/Ag膜渗透系统进行氚的回收.     

Advantages of high-field tokamaks for fusion reactor development

High-field designs could reduce the cost and complexity of tokamak reactors. Moreover, the certainty of achieving required plasma performance could be increased. Strong Ohmic heating could eliminate or significantly decrease auxiliary heating power requirements and high values of n_E could be obtained in modest-size plasmas. Other potential advantages are reactor operation at modest values of , capability of higher power density and wall loading, and possibility of operation with advanced fuel mixtures. Present experimental results and basic scaling relations imply that the parameterB ²a, where B is the magnetic field and a is the minor radius, may be of special importance. A superhigh-field compact ignition experiment with very high values ofB ²a (e.g.,B ²a=150 T² m) has the potential of Ohmically heating to ignition. This short-pulse device would use inertially cooled copper plate magnets. Compact engineering test reactor and/or experimental hybrid reactor designs would use steady-state, water-cooled copper magnets and provide long-pulse operation. Design concepts are also described for demonstration/commercial reactors. These devices could use high-field superconducting magnets with 7–10 T at the plasma axis.