Fission-suppressed blankets for fissile fuel breeding fusion reactors

Two blanket concepts for deuterium-tritium (DT) fusion reactors are presented which maximize fissile fuel production while at the same time suppress fission reactions. By suppressing fission reactions, the reactor will be less hazardous, and therefore easier to design, develop, and license. A fusion breeder operating a given nuclear power level can produce much more fissile fuel by suppressing fission reactions. The two blankets described use beryllium for neutron multiplication. One blanket uses two separate circulating molten salts: one salt for tritium breeding and the other salt for U-233 breeding. The other uses separate solid forms of lithium and thorium for breeding and helium for cooling.Nuclear power is the sum of fusion (D + T 14 MeV neutron+ 3.5 MeV alpha) power plus additional power from neutron-induced reactions in the blanket.     

End product economics and fusion research program priorities

It is shown that deuterium based fusion fuels and reactors based on them face severe technological disadvantages in comparison with fission based systems as power sources for central station electric power plants. The author postulates the most plausible deuterium based fusion reactor consistent with the physics of the fusion reaction itself and compares this reactor (called OMR-DT) with existing fission reactors. Since neutrons are the main problem in fusion, the author suggests that a great deal more effort should be given to the study of non-Maxwellian plasmas with the emphasis on neutron-free fuel cycles. The author also suggests that the deuterium based fusion driver may play its best role as a fissile fuel producer.     

Optimized fissile and fusile breeding in a laser-fusion fissile-enrichment flux trap blanket

Optimization of fissile and fusile production in the SOLASE-H laser-fusion fissile-enrichment fuel-factory blanket is carried out. The objective is maximizing fissile breeding with the constraints of maintaining self-sufficiency in tritium production, and realistically accounting in the modeling for structural and coolant compositions and configurations imposed by the thermal-hydraulic and mechanical designs. The effect of radial and axial blanket zone thicknesses on fusile and fissile breeding is studied using a procedure which modifies the zones' effective optical thicknesses, rather than the actual three-dimensional geometrical configurations. A tritium yield per source neutron of 1.08 and a Th (n, ) reaction yield per source neutron of 0.43 can be obtained in such a concept, where ThO₂ Zircaloy-clad fuel assemblies for light water reactors (LWRs) are enriched in the²³³U isotope by irradiating them in a lead flux trap. This corresponds to 0.77 kg/[MW(th)-year] of fissile fuel production, and 1.94 years of irradiation in the fusion reactor to attain an average 3 w/o fissile enrichment in the fuel assemblies. For a once-through LWR cycle, a support ratio of 2–3 is estimated. However, with fuel recycling, more attractive support ratios of 4–6 may be attainable for a conversion ratio of 0.55, and of 5–8 for a conversion ratio of 0.70. These estimates are lower than those reported, around 20, for related designs.     

Radiotoxicity and Residual Energy Release of Actinides and Fission Products of Spent VVÉR Fuel in Long-Term Storage

The time dependence of the residue energy release and radiotoxicity of spent VVÉR-1000 nuclear fuel with long-term storage or uniform accumulation in long-term storage is investigated. The calculations of the energy release take account of the contribution of , , and radiation, and the calculations of the radiotoxicity take account of the maximum admissable activity of nuclides in water and air. The data presented can be used for developing a strategy for long-term storage of spent nuclear fuel from power reactors. 4 figures, 4 tables, 4 references.     

加速增产核燃料的天然安全“核热泉”快中子增殖堆

为保证21世纪中国经济的持续稳定地高速增长,必须充分发挥核能的巨大潜力,使之配合其他可再生能源同步增长,及早大规模替代煤炭等化石能源。由于目前国内大量兴建的核电站以压水堆为主,需要消费大量天然铀资源,倚靠廉价铀供应难于维持长期增长,必须依靠快中子增殖生产人造裂变燃料——钚,才能摆脱天然铀原料短缺的束缚。然而,传统的快中子增殖堆的核燃料增产速度较慢,难于配合中国核电的高速增长。本文介绍一种先进快中子增殖堆(AFBR)方案,其中利用在线连续换料的空心球形燃料元件,依靠载热剂的出入口之间的温度差实现满功率自然循环,可以成倍地提高燃料比功率与核燃料增殖速度。本快中子增殖堆改进了俄罗斯称为"天然安全"的BREST铅冷快堆设计方案,成为无须人为控制的"核热泉",它能在不设置加压泵及高位铅池的情况下,自动按外部负荷需要供应必要的热量,完全依靠自然循环将全部裂变热能及停堆后堆芯余热散出,不至对环境产生放射性污染。     

Nuclear fusion from crack-generated particle acceleration

Summary In summary, the high-voltages necessary to accelerate deuterons to energies sufficient to produce modest numbers (10⁴–10⁵/sec) of d-d neutrons appears to be possible as a result of cracking or fracture of the metal lattice in the cold fusion experiments.This mechanism requires neither massive electrons nor exotic nuclear reactions to explain the apparent cold fusion d-d neutron production results. Instead, it is possible that high voltage electrostatic fields, known to be associated with cracking, can reside across a crack gap long enough for the deuterons to be accelerated to sufficiently high energy to produce the d-d reactions. Interestingly, the electrostatic acceleration is quite similar to that of laboratory accelerators except for its submicron scale. Clearly, much work is still required to determine whether such a crack-generated acceleration mechanism, a quasi-particle mechanism, some combination of these, or some other, as yet unidentified mechanism is responsible for the nuclear effects seen in cold fusion experiments.Presented at the Workshop on Cold Fusion Phenomena, Sante Fe, New Mexico, May 23–25, 1989.     

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%.     

Graphite-moderated,gas-cooled,and water-moderated,water-cooled reactors as power units in nuclearelectric power stations

The present article reviews a number of papers submitted at the Second International Conference on the Peaceful Uses of Atomic Energy bearing on water-cooled, water-moderated, graphite-moderated, and gas-cooled reactors abroad.The basic characteristics of all of the operational power reactors, as well as of the high-power, graphite-gas water-cooled, and water-moderated reactors built abroad are cited in the article.Differences in the pathways and means of development of nuclear power in different countries are given due acknowledgement. In Britain, for example, after the first generation reactors (Calder Hall type) were built, advanced second generation reactors (Hinkley Point, etc.) were introduced, and intensive studies are now underway on third generation reactors. The ultimate purpose of the research in progress is the development of a high- temperature reactor operating in a unit with the associated gas turbine. A transition in the USA from the classical pressurized-water reactor concept (PWR) to that of the boiling water reactor with direct steam feed to the turbine is planned.The continuous improvement in the efficiency of graphite-moderated, gas-cooled reactors and water-cooled, water-moderated reactors and the reduction in capital costs per unit rated power and in cost of power developed is shown.     

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.     

Suitability of Small Scale Linear Systems for a Fission–Fusion Reactor, Breeder, and Waste Transmutation

To date the magnetic fusion effort has been almost entirely devoted to only one application, that being a multi gigawatt central station electric plant. Given the already well established fission based industry, the likelihood that fusion will have any impact on curbing the current carbon-based energy demand in the 21st century is slim. This paper advocates that the first and primary use of fusion neutrons should be as the driver for a sub-critical fission nuclear energy system—a fission–fusion hybrid reactor. This system can also be utilized to transmute long-lived radioactive wastes, and breed fissile nuclear fuel for several additional fission reactors. A small-scale fusion system based on a reciprocating fusion cycle employing the magneto-kinetic compression of the Field Reversed Configuration (FRC) is particularly well suited for this application. The characteristics of this fusion neutron driver and the potential for transmutation of long-lived nuclear wastes and breeding of fissile nuclear fuel in a blanket are presented.     

The reversed-field-pinch (RFP) fusion neutron source: A conceptual design

The conceptual design of an ohmically heated, reversed-field pinch (RFP) operating at 5-MW/m² steady-state DT fusion neutron wall loading and 124-MW total fusion power is presented. These results are useful in projecting the development of a cost effective, low-input-power (206 MW) source of DT neutrons for large-volume (10 m³), high-fluence (3.4 MW yr/m²) fusion nuclear materials and technology testing.Work supported by U.S. DOE.     

The Potential for Developing and the Possibility of Attaining Radiation Equivalence between Uranium and Wastes in Future Nuclear Power

The development of nuclear power with capacity up to 300 GW using thermal reactors and BREST fast reactors with small excess breeding (BR 1.05) and nuclear power operating potentially for up to 3750 yr using depleted uranium is examined on the basis of fuel materials balances. To examine the radiation balance, all wastes, which have accumulated by the time mentioned from reprocessing of spent fuel from thermal and fast reactors are included, taking account of the running nuclide composition. The change in the potential biological danger is calculated as a function of the holding time of the entire mass of the wastes taking account of the short-lived nuclide daughter products. The possibility of starting coextraction of thorium and radium with uranium starting in 2010 and 2030 or without coextraction is taken into account. If coextraction is implemented during the periods indicated, then radiation balance of the radiactive wastes which accumulate by 2100 or 2200 is reached within a holding period of 80–120 yr. Without coextraction, the fraction of plutonium going into the wastes will have to be decreased from 0.1 to 0.01%.     

聚变—裂变混合堆及其在我国核能发展中的作用

本文概要介绍聚变和聚变-裂变混合堆基本原理及其作用。聚变-裂变混合堆可以为压水堆或快堆提供充足的核燃料。它和压水堆或快堆组成的系统具有经济可行性。在解决我国核能发展中燃料短缺问题和促进纯聚变能源的发展方面可望发挥重要的作用。     

Criticality theory for a reactor in the case of fluctuations in the fuel density

Conclusions In Fig. 2 we show graphs of the dependence of the additional reactivity that arises as a result of fluctuations of the fuel density. As follows from Fig. 2, the increase in the reactivity for sufficiently large reactors and for (/₀) 0.1–0.05 is comparable with the contribution of the delayed neutrons. Thus, it is in principle possible to regulate the criticality of the reactor by exciting fluctuations of the density of a gaseous fuel. Regulation in this manner has decided advantages. Thus, the time in which the reactivity changes which determines the transient processes, is short — of the order of one period of the fluctuation. Moreover, there is practically no danger of accidents, since the reactivity falls as soon as the fluctuations cease. The amplitude of the fluctuation of the neutron flux (see, for example, the expression (20)) always exceeds the amplitude of the fluctuations in the fuel density by (k–1)^–1/2. This circumstance may be exploited to obtain a neutron flux pulsating with a large amplitude.This effect of a growth in thereactivity as a result of fluctuations of the fuel density may prove important in the study of the possibility of self-oscillatory conditions of operation in reactors with a high neutron flux.Translated from Atomnaya Énergiya, Vol. 27, No. 2, pp. 107–111, August, 1969.     

Neutronic analysis of a thorium fusion breeder with enhanced protection against nuclear weapon proliferation

The fissile breeding capability of a (D,T) fusion-fission (hybrid) reactor fueled with thorium is analyzed to provide nuclear fuel for light water reactors (LWRs). Three different fertile material compositions are investigated for fissile fuel breeding: (1) ThO₂; (2) ThO₂ denaturated with 10% natural-UO₂ and (3) ThO₂ denaturated with 10% LWR spent fuel. Two different coolants (pressurized helium and Flibe ‘Li₂BeF₄’) are selected for the nuclear heat transfer out of the fissile fuel breeding zone. Depending on the type of the coolant in the fission zone, fusion power plant operation periods between 30 and 48 months are evaluated to achieve a fissile fuel enrichment quality between 3 and 4%, under a first-wall fusion neutron energy load of 5 MW/m² and a plant factor of 75%. Flibe coolant is superior to helium with regard to fissile fuel breeding. During a plant operation over four years, enrichment grades between 3.0 and 5.8% are calculated for different fertile fuel and coolant compositions. Fusion breeder with ThO₂ produces weapon grade ²³³U. The denaturation of the ²³³U fuel is realized with a homogenous mixture of 90% ThO₂ with 10% natural-UO₂ as well as with 10% LWR spent nuclear fuel. The homogenous mixture of 90% ThO₂ with 10% natural-UO₂ can successfully denaturate ²³³U with ²³⁸U. The uranium component of the mixture remains denaturated over the entire plant operation period of 48 months. However, at the early stages of plant operation, the generated plutonium component is of weapon grade quality. The plutonium component can be denaturated after a plant operation period of 24 and 30 months in Flibe cooled and helium cooled blankets, respectively. On the other hand, the homogenous mixture of 90% ThO₂ with 10% LWR spent nuclear fuel remains non-prolific over the entire period for both, uranium and plutonium components. This is an important factor with regard to international safeguarding.     

D-T experiments on TFTR

Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, alpha confinement, alpha heating and possible alpha driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of 20 MW of tritium and 14 MW of deuterium neutral beams into the TFTR produced a plasma with a T/D density ratio of 1 and yielded a maximum fusion power of 9.2 MW. The fusion power density in the core of the plasma was 1.8 MW m^–3 approximating that expected in a D-T fusion reactor. In other experiments TFTR has produced 6.4 MJ of fusion energy in one pulse satisfying the original 1976 goal of producing 1 to 10 MJ of fusion energy per pulse. A TFTR plasma with T/D density ratio of 1 was found to have 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass, A, of _E. The core ion temperature increased from 30 keV to 37 keV due to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 keV to 10.6 keV can be attributed to electron heating by the alpha particles. At fusion power levels of 7.5 MW, fluctuations at the Toroidal Alfvén Eigenmode frequency were observed by the fluctuation diagnostics. However, no additional alpha loss due to the fluctuations was observed. These D-T experiments will continue over a broader range of parameters and higher power levels.Work supported by U.S. Department of Energy Contract No. DE-AC02-76-CHO-3073.     

The case for the fusion hybrid

The use of nuclear fusion to produce fuel for nuclear fission power stations is discussed in the context of a crucial need for future energy options. The fusion hybrid is first considered as an element in the future of nuclear fission power to provide long term assurance of adequate fuel supplies for both breeder and convertor reactors. Generic differences in neutronic characteristics lead to a fuel production potential of fusion-fission hybrid systems which is significantly greater than that obtainable with fission systems alone. Furthermore, cost benefit studies show a variety of scenarios in which the hybrid offers sufficient potential to justify development costs ranging in the tens of billions of dollars. The hybrid is then considered as an element in the ultimate development of fusion electric power. The hybrid offers a near term application of fusion where experience with the requisite technologies can be derived as a vital step in mapping a credible route to eventual commercial feasibility of pure fusion systems. Finally, the criteria for assessment of future energy options are discussed with prime emphasis on the need for rational comparison of alternatives. This approach is contrasted with the dual standard too often used in judging the risks and benefits of nuclear power where, for example, rather minor radiological effects are highlighted while much larger exposures to radiation from medical x-rays, airplane travel, color television sets, etc., are ignored. It is concluded that the fusion hybrid deserves a prominent place among new energy resources but that early attention to insure an adequately informed public is a vital ingredient in assuring reasonable prospects of success.     

Circulation loop of a nuclear reactor as a source of radiation,particularly for radiation chemistry

At the present time it is necessary to solve the problems concerned with the use of powerful radiation fluxes from nuclear reactors for irradiation in various scientific and technological fields, and, in particular, for carrying out radiation-chemical processes. In this work we discuss some of the aspects of employing a reactor for these purposes. We examine a circulation loop which transfers -radiation from a reactor to a radiation apparatus for those cases where the activated material forms one radioisotope, without radioactive daughters. The theory is reduced to working formulas and graphs which make it possible to calculate the strength of the -radiation emitted in the apparatus and to choose the parameters which would ensure the most effective use of the loop.     

Commercial objectives,technology transfer,and systems analysis for fusion power development

Fusion is an essentially inexhaustible source of energy that has the potential for economically attractive commercial applications with excellent safety and environmental characteristics. The primary focus for the fusion-energy development program is the generation of centralstation electricity. Fusion has the potential, however, for many other applications. The fact that a large fraction of the energy released in a DT fusion reaction is carried by high-energy neutrons suggests potentially unique applications. These include breeding of fissile fuels, production of hydrogen and other chemical products, transmutation or burning of various nuclear or chemical wastes, radiation processing of materials, production of radioisotopes, food preservation, medical diagnosis and medical treatment, and space power and space propulsion. In addition, fusion R&D will lead to new products and new markets.Each fusion application must meet certain standards of economic and safety and environmental attractiveness. For this reason, economics on the one hand, and safety and environment and licensing on the other hand, are the two primary criteria for setting long-range commercial fusion objectives. A major function of systems analysis is to evaluate the potential of fusion against these objectives and to help guide the fusion R&D program toward practical applications. The transfer of fusion technology and skills from the national laboratories and universities to industry is the key to achieving the long-range objective of commercial fusion applications.