The zero-waste option for nuclear fusion reactors: Advanced fuel cycles and clearance of radioactive materials

This paper deals with the radioactive waste issue for fusion reactors, proposing an innovative solution (the “zero-waste” option), which could be a clear advantage of fusion power versus fission, in view of its ultimate safety and public acceptance. Its goal is the Clearance (declassification to non-radioactive materials) of all reactor components, after a sufficient period of interim decay, according to the recommendations recently issued by IAEA.Even if feasible in theory, a zero-waste option for fusion reactors using the Deuterium–Tritium fuel cycle will be difficult to obtain in practice: a relevant amount of radioactive materials from reactor decommissioning – even if recyclable within the nuclear industry – should be disposed of as low-level waste.As a further step towards the zero-waste option, the features of fusion reactors based on alternative advanced fuel cycles have been examined, to assess whether that goal could be reached for such devices. Fusion reactors with advanced Deuterium–Helium-3 (DHe) fuel cycle have quite outstanding environmental advantages. Compact ignition tokamaks can be designed in order to achieve DHe ignition. Ignitor, a compact ignition experiment aimed at studying DT plasmas, may also be used in that direction. The extrapolation of Ignitor technologies towards a larger and more powerful experiment using advanced fuel cycles (Candor) is described. Results show that Candor does reach the zero-waste option.A fusion power reactor based on the DHe cycle could be the ultimate correct response to the environmental requirements for future nuclear power plants.     

A low wall-loading DEMO reactor design with high priority for early and reliable realization of a tokamak fusion reactor over the cost performance

Based on scientific databases adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R&D program, a low wall-loading DEMO fusion reactor has been designed, where high priority has been given to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the major radius of this DEMO reactor is chosen to be 10 m, plasma ignition is achievable with a low fusion power of 0.8 GW and an operation period of 4–5 hours is available only with inductive current drive. The low ignition power makes it possible to adopt a first wall with an austenitic stainless steel, for which significant databases and operating experience exists, due to its use in the presence of neutron irradiation in fission reactors. In step with development of advanced materials, a step-wise increase of the fusion power seems to be feasible and realistic, because this DEMO reactor has the potential to produce a fusion power of 5 GW.     

A 2-D model of plasma-jet interaction in the hylife inertial confinement fusion reactor

A two-dimensional, axisymmetric model was developed to study the impulsive crossflow of a lithium plasma through a close-packed annular arrangement of liquid jets, a problem that arises in the design of a certain class of inertial confinement fusion reactors. Calculations performed with this model indicate that a 5 m radius reactor cavity, standing 8 m high, can deliver 1 GWe when pulsed at 1 Hz, without adverse loading of the first wall.     

Compact Stellarators

Stellarators offer advantages for reactors, namely the potential for steady state operation with low recirculating power (high engineering Q) and without disruptions. A substantial portion of the world fusion program is devoted to the development of stellarators as a magnetic confinement system. The world stellarator program, as it currently exists, is focused on high-aspect-ratio (R/a = 5 ? 11) designs that lead to very large reactors. For example the German advanced stellarator reactor design HSR has an aspect ratio of 12 and a major radius of 22 m. An important issue for stellarator research is whether more compact reactor designs are possible. Could the advantage of stellarators also be realized at dimensions and performance levels closer to those of the advanced tokamak reactor ARIES-RS (R = 5.5 m, neutron wall load of 4 MW/m2)? Theory has identified a class of “compact stellarator” plasma configurations that could be the basis for such a design. They are promising, but need to be studied experimentally in order to realistically assess their potential. The most cost-effective way to accomplish this is to carry out the compact stellarator proof-of-principle program that has been proposed by the U.S. stellarator community. This program would answer the basic physics questions for compact stellarators and make important contributions to the world stellarator knowledge base at a cost (about $30M/year) that is modest compared to expenditures for stellarator and tokamak research world-wide.     

The reversed field pinch reactor

The physical properties and engineering concepts of reversed field pinch reactors are presented and illustrated by two recent designs developed at Culham Laboratory and Los Alamos Scientific Laboratory. Both designs employ ohmic heating to ignite the plasma, operate in the pulsed mode without refuelling during the burn, and utilize superconducting magnetic field coils. Similarities between the two designs include the choice of plasma minor radius of 1.5 m, the use of low wall loading and electrical power output under 1 GW, and reasonable agreement on plasma behaviour. Major differences occur in the engineering design. The Culham design includes a separate radiation shield first wall behind which lies a passive stabilising shell. Large segments are required, implying the movement of toroidal field coils for maintenance of the helium cooled blanket. In the LASL design more conventional pressurized water cooling technology is used and the radiation shield and passive shell are combined, sacrificing thermal efficiency for an improvement in maintainability. The advantages of the reversed field pinch in higher beta, lower toroidal field, lower magnet costs and lack of auxiliary heating requirements appear at this stage to be significant when comparison is made with a pulsed tokamak reactor designed under the same ground rules.     

A Review on the Potential Use of Austenitic Stainless Steels in Nuclear Fusion Reactors

Various engineering materials; austenitic stainless steels, ferritic/martensitic steels, vanadium alloys, refractory metals and composites have been suggested as candidate structural materials for nuclear fusion reactors. Among these structural materials, austenitic steels have an advantage of extensive technological database and lower cost compared to other non-ferrous candidates. Furthermore, they have also advantages of very good mechanical properties and fission operation experience. Moreover, modified austenitic stainless (Ni and Mo free) have relatively low residual radioactivity. Nevertheless, they can’t withstand high neutron wall load which is required to get high power density in fusion reactors. On the other hand, a protective flowing liquid wall between plasma and solid first wall in these reactors can eliminate this restriction. This study presents an overview of austenitic stainless steels considered to be used in fusion reactors.     

Application of Galerkin Method to Hydraulic Analysis for Turbulent Fluid Flow around LMFBR Fuel Rod

Wide range of parameter surveys are made on the DT fusion tokamak experimental reactor next to JT-60. Various physics and engineering requirements are taken into account, e.g. self-ignition, available maximum toroidal β value, α-particle confinement, total fusion power, neutron wall loading, heat flux to divertor plate, structural restriction on major radius, device size, maximum toroidal magnetic field, poloidal field power supply and so on. Theoretical scaling law for the available maximum toroidal β value determined by ballooning mode instability is used. The toroidal magnetic field on plasma axis can be expressed by the aspect ratio A for a given maximum field at the toroidal field coil conductor. Empirical scaling law for the electron energy confinement and neoclassical heat conductivity for the ion are employed. These confinement times can be expressed by the plasma minor radius a and A through the maximum available β value and the toroidal field on axis. In the similar way, most of the physics and engineering requirements can be mapped on the a-A diagram. This diagram enables us to make systematic and wide range of parameter surveys of the device. In particular, this offers a clear perspective on the device parameters, which can mitigate the engineering difficulties and can also realize the required plasma performances.     

Prompt and power reactivity coefficients for the next generation VVER-1000 reactor including hexagonal assemblies and annular fuels

Many neutronics as well as thermal-hydraulics calculations have been made to find the performance of the proposed annular fuels (internally and externally cooled fuel pins) for both next generation PWRs and BWRs. Specifically, there has not been a significant study on the Russian type VVER-1000 reactors with annular fuels. Our aim herein is to study two important safety coefficients of the Iranian VVER-1000 core including hexagonal annular fuel assemblies at its BOC. The safety coefficients are “prompt reactivity coefficient” and “power reactivity coefficient”, where all simulations are made using MCNP-5 code. We found less (absolutely) Doppler coefficient for the next generation VVER-1000 and therefore Doppler coefficient decreasing is a good feature to avoid more resonance neutrons absorbing in the U-238; causes more fission density and also less soluble boron for core controlling (at the BOC) with comparing to the current VVER-1000 solid pins.     

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.     

Continuous tokamak operation with an internal transformer

A large improvement in efficiency of current drive in a tokamak can be obtained using neutral beam injection to drive the current in a plasma which has low density and high resistivity. The current established under such conditions acts as the primary of a transformer to drive current in an ignited high-density plasma. In the context of a model of plasma confinement and fusion reactor costs, it is shown that such transformer action has substantial advantages over strict steady-state current drive. It is also shown that cycling plasma density and fusion power is essential for effective operation of an internal transformer cycle. Fusion power loading must be periodically reduced for intervals whose duration is comparable to the maximum of the particle confinement and thermal inertia time scales for plasma fueling and heating. The design of neutron absorption blankets which can tolerate reduced power loading for such short intervals is identified as a critical problem in the design of fusion power reactors.     

The fusion breeder concept

Some practical aspects and broader consequences of strategies that have been proposed for the association of fusion and fission processes are reviewed.It is concluded more particularly that the build-up of the technology necessary for the ultimately desired substitution of fusion for the present fission source of nuclear power could be much advanced by an interim period of the application of fusion entirely to the breeding of U-233 for the large number of fission reactors that will be coexisting during this period.Principal reasons for this view are the considerable technical easements of this approach, and the very large gain in effective energy release per fusion event, which offers economic advantage long before achievement of the classic definition of the fusion “break even” point.     

聚变-裂变混合堆水冷包层中子物理性能研究

研究直接应用国际热核聚变实验堆（ITER）规模的聚变堆作为中子驱动源,采用天然铀为初装核燃料,并采用现有压水堆核电厂成熟的轻水慢化和冷却技术,设计聚变-裂变混合堆裂变及产氚包层的技术可行性。应用MCNP与Origen2相耦合的程序进行计算分析,研究不同核燃料对包层有效增殖系数、氚增殖比、能量放大系数和外中子源效率等中子物理性能的影响。计算分析结果显示,现有核电厂广泛使用的UO₂核燃料以及下一代裂变堆推荐采用的UC、UN和U₉₀Zr₁₀等高性能陶瓷及合金核燃料作为水冷包层的核燃料,都能满足以产能发电为设计目标的新型聚变 裂变混合堆能量放大倍数的设计要求,但只有UC和U₉₀Zr₁₀燃料同时满足聚变燃料氚的生产与消耗自持的要求。研究结果对进一步研发满足未来核能可持续发展的新型聚变-裂变混合堆技术具有潜在参考价值。     

Yield from Proton-Induced Reaction on Light Element Isotopes in the Hydrogen Plasma Focus

The high Q-value of some (p,α) fusion reactions is very important in the investigation that can lead to power production with controlled fusion using advanced fuels (hydrogen-lithium-7, hydrogen-boron-11). For this reason, it is crucial to know the rates of these fusion reactions. Unfortunately, in the fusion machines such as plasma focus device, the interaction energy is usually far below the Coulomb barrier. Because of that, direct measurements of the relevant reaction cross sections are practically impossible. A few different indirect approaches have been proposed. In this work the Trojan Horse Method (THM) will be described. On the basis of the results obtained from the THM method and data, which are well-known from our previous work (Banjanac et al. in Radiat Meas 40:483–485, 2005), the reaction rate for proton-induced reaction ⁷Li(p,α)α produced in the hydrogen plasma focus is calculated. This calculation will be compared with the measurements of α particles production rate using CR-39 detectors.     

Design and fabrication of a high-temperature helium regenerator

Refractory metallic foams can increase heat transfer efficiency in gas-to-gas and liquid metal-to-gas heat exchangers by providing an extended surface area for better convection, i.e. conduction into the foam ligaments providing a “fin-effect,” and by disruption of the thermal boundary layer near the hot wall and ligaments by turbulence promotion. In this article, we describe the design of a high-temperature refractory regenerator (closed-loop recuperator) using computational fluid dynamics (CFD) modeling of actual foam geometries obtained through computerized micro-tomography. The article outlines the design procedure from geometry import through meshing and thermo-mechanical analysis and discusses the challenges of fabrication using pure molybdenum and TZM. The foam core regenerator is more easily fabricated, less expensive and performs better than refractory flat plate-type heat exchangers. The regenerator can operate with a maximum hot leg inlet temperature of 900 °C and transfer 180 kW to the cold leg using 100 g/s helium at 4 MPa. Future high heat flux experiments on helium-cooled plasma facing components will utilize the high temperature and high pressure capabilities of this unique regenerator. Similar components will be required to adapt fusion power reactors to high-efficiency Brayton power conversion systems and enable operation of advanced divertor and blanket systems.     

Optimization of nonthermal fusion power consistent with channeling of charged fusion product energy

If the energy of charged fusion products can be diverted directly to fuel ions, non-Maxwellian fuel ion distributions and temperature differences between species will result. To determine the importance of these nonthermal effects, the fusion power density is optimized at constant- for nonthermal distributions that are self-consistently maintained by channeling of energy from charged fusion products. For D-T and D-³He reactors, with 75% of charged fusion product power diverted to fuel ions, temperature differences between electrons and ions increase the reactivity by 40–70%, while non-Maxwellian fuel ion distributions and temperature differences between ionic species increase the reactivity by an additional 3–15%.     

Status and prospect of the development of liquid lithium limiters for stellarotor TJ-II

Stellarator concept is considered as a promising approach for power fusion reactor development because it basically free from disruptions and other extreme thermal load events. However, the potential problem of impurity accumulation in stellarator plasma should be taken into account. Very promising results in density control, plasma reproducibility and confinement characteristics have been obtained with application of “lithiation” technology. The next step in the improvement of TJ-II Heliac plasma performance is the development and creation of two poloidal liquid lithium limiters (LL). Experimental possibilities, design, structural materials and main parameters of LL based on capillary-pore structure (CPS) filled with liquid lithium are considered. Understanding in hydrogen isotope interaction with liquid lithium surface is an important aspect of lithium technology development for fusion reactor application. Therefore study of deuterium sorption/desorption process on a lithium surface of LL is stipulated. The development of lithium CPS based devices decreasing intensity of plasma–wall interaction on a central “groove” of TJ-II vacuum camera is proposed as the further step in plasma performance improvement owing to decrease in impurity flux from the wall.     

Some comments on energy and particle confinement in advanced fuel EBT reactors

Results of a point model calculation for advanced fuel (cat. D and D³He) EBT reactors are used to determine some of the limitations on the ratio of ion particle to energy confinement time. The greater fraction of charged fusion products produced in the advanced fuel reactions and the greater fraction of their energy radiated cause the effect of on ash buildup to be a factor of 4 greater for the advanced fuels than that of DT fuel. Hence it is found that<5 for steady state ignited advanced fuel EBT reactors, whereas 22 is the restriction for DT fueled EBT reactors. A survey of for neoclassical bumpy torus ions reveals that in the plateau regime,<5 appears possible but is critically dependent on the nature of the electric field.     

A Review of Confinement Requirements for Advanced Fuels

The energy confinement requirements for burning D-³He, D-D, or P-¹¹B are reviewed, with particular attention to the effects of helium ash accumulation. It is concluded that the DT cycle will lead to the more compact and economic fusion power reactor. The substantially less demanding requirements for ignition in DT (the n_{e E} T required for ignition in DT is smaller than that of the nearest advanced fuel, D-³He, by a factor of 50) will allow ignition, or significant fusion gain, in a smaller device; while the higher fusion power density (the fusion power density in DT is higher than that of D-³He by a factor of 100 at the same plasma pressure) allows for a more compact and economic device at fixed fusion power.     

Utilization of Ceramic Uranium Fuels in ARIES-RS Fusion Reactor

This study presents the neutronic performance of the ARIES-RS fusion reactor design using different natural ceramic uranium fuels, namely UO₂, UN or U₃Si₂, dispersed in graphite matrix. These fissionable fuels inserted as micro spheres into the first range quadratic channels at the immediate neighborhood of the first wall in the inboard blanket to amplify fusion power and breed fissile fuel. Neutron transport calculations were performed with the help of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S₈–P₃ approximation. Among the investigated fuels, UN showed the best neutronic performance while UO₂ and U₃Si₂ had similar performances. Numerical results pointed out that inserting fissionable fuel zone even with a small thickness (10 cm) in a pure fusion reactor increased fusion power from 2170 MW to 4500, 5250 and 4150 MW depending on the fuel type. Furthermore significant amount of fissile fuel was produced to be charged to light water reactors.     

Simulation of a gas-cooled fluidized bed nuclear reactor – Part II: Stability of a fluidized bed reactor with mixed oxide fuels

Fertile materials can be converted by nuclear reaction into a transuranic mixture with a fissile content, mostly plutonium isotopes. Fuel which includes a limited proportion of plutonium is already used in some reactors. Use is restricted by the smaller delayed neutron yield and lower negative temperature coefficient of reactivity compared with uranium fuels.