Fusion-fission hybrid design with analysis of direct enrichment and non-proliferation features (the SOLASE-H study)

The role of a fusion-fission hybrid in the context of a nuclear economy with and without reprocessing is examined. An inertial confinement fusion driver is assumed and a consistent set of reactor parameters are developed. The form of the driver is not critical, however, to the general concepts. The use of the hybrid as a fuel factory within a secured fuel production and reprocessing center is considered. Either the hybrid or a low power fission reactor can be used to mildly irradiate fuel prior to shipment to offsite reactors thereby rendering the fuel resistant to diversion. A simplified economic analysis indicates a hybrid providing fuel to 10 fission reactors of equal thermal power is insensitive to the recirculating power fraction provided reprocessing is permitted. If reprocessing is not allowed, the hybrid can be used to directly enrich light water reactor fuel bundles fabricated initially from fertile fuel (either ThO₂ or ²³⁸UO₂). A detailed neutronic analysis indicates such direct enrichment is feasible but the support ratio for ²³³U or ²³⁹Pu production is only 2, making such an approach highly sensitive to the hybrid cost. The hybrid would have to produce considerable net power for economic feasibility in this case. Inertial confinement fusion performance requirements for hybrid application are also examined and an integrated design, SOLASE-H, is described based upon the direct enrichment concept.     

Reprocessing free nuclear fuel production via fusion fission hybrids

Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively “new” cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th²³²–U²³³ conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO₂ matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U²³⁵ fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.     

Development of advanced HTR fuel elements

Spherical fuel elements with coated particles are the essential constituents of the integrity of a pebble-bed reactor, particularly with respect to fission product retention. While all gas-cooled reactors exhibit an extremely low primary circuit activity, recent developments have reduced the source term to virtually zero. This entails an enormous effort in the characterization of the fuel behaviour in manufacture, during irradiation, and in accident testing.With modern SiC-coated particles it can be shown that all radiologically relevant fission products are completely retained inside the silicon carbide layer of intact fuel particles. The dominant source term for fission product release will, therefore, be the small number of particles with defective coatings. Permanent damage to the fission product retaining SiC layer can only start in accidents involving temperatures significantly beyond 1600°C.     

Radionuclide inventories in the discharged fuels of PHWR-220, BWR-160, VVER-1000 and the conceptual ATBR-600 reactors – A case study

Radionuclides content in the discharged fuel of the conceptual thorium breeder reactor ATBR-600 has been assessed and compared against other thermal power reactors considered in Indian nuclear power programme. The contribution of actinides and the fission products inventories in the discharged fuels are separately estimated and assessed. The ATBR-600 reactor is suggested for closed fuel cycle option. The relatively large presence of the unspent plutonium would in fact be recycled. Nonetheless, the data has been presented in the event of operating ATBR-600 like other present day power reactors in a once through fuel cycle mode.     

On the safety of conceptual fusion-fission hybrid reactors

A preliminary examination of some potential safety questions for conceptual fusion-fission hybrid reactors is presented in this paper. The study and subsequent analysis was largely based upon one design, a conceptual mirror fusion-fission reactor, operating on the deuterium-tritium plasma fusion fuel cycle and the uranium-plutonium fission fuel cycle. The major potential hazards were found to be: (a) fission products, (b) actinide elements, (c) induced radioactivity, and (d) tritium.As a result of these studies, it appears that highly reliable and even redundant decay heat removal must be provided. Loss of the ability to remove decay heat results in melting of fuel, with ultimate release of fission products and actinides to the containment. In addition, the studies indicate that blankets can be designed which will remain subcritical under extensive changes in both composition and geometry. Magnet safety and the effects of magnetic fields on thermal parameters were also considered.     

The fusion breeder

The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the U.S. fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the U.S. fusion program and the U.S. nuclear energy program. There is wide agreement that many approaches will work and will produce fuel for five equal-sized LWRs, and some approach as many as 20 LWRs at electricity costs within 20% of those at today's price of uranium ($30/lb of U₃O₈). The blankets designed to suppress fissioning, called symbiotes, fusion fuel factories, or just fusion breeders, will have safety characteristics more like pure fusion reactors and will support as many as 15 equal power LWRs. The blankets designed to maximize fast fission of fertile material will have safety characteristics more like fission reactors and will support 5 LWRs. This author strongly recommends development of the fission suppressed blanket type, a point of view not agreed upon by everyone. There is, however, wide agreement that, to meet the market price for uranium which would result in LWR electricity within 20% of today's cost with either blanket type, fusion components can cost severalfold more than would be allowed for pure fusion to meet the goal of making electricity alone at 20% over today's fission costs. Also widely agreed is that the critical-path-item for the fusion breeder is fusion development itself; however, development of fusion breeder specific items (blankets, fuel cycle) should be started now in order to have the fusion breeder by the time the rise in uranium prices forces other more costly choices.     

Design and analysis on tritium system of multi-functional experimental fusion–fission hybrid reactor (FDS-MFX)

As early application of fusion technology, the fusion–fission hybrid systems/reactors could be used to transmute long-lived radioactive waste and produce fissile nuclear fuel. A fusion–fission hybrid reactor named FDS-MFX was designated for checking and validating the DEMO reactor blanket relevant technologies. The reactor design is based on easy-achieved plasma parameters extrapolated from the successful operation of tokamaks and the subcritical blanket is designed based on the well-developed technologies of fission reactors. In this contribution, a concept of the tritium system was designed for the FDS-MFX: the tritium was extracted from LiPb by the helium purge gas which contains a small amount of hydrogen gas, then the impurity gas was removed by cold trap, finally tritium was separated from hydrogen isotope by the cryogenic distillation and supply to reactor core. On the basis of data obtained by present design and experimental research, the system parameters were presented and discussed in detail. The results preliminarily demonstrated the engineering feasibility of the design.     

Regulatory aspects of fusion power-lessons from fission plants

Experience from fission reactors has shown the regulatory process for licensing a nuclear facility to be legalistic, lengthy, unpredictable, and costly. This experience also indicates that much of the regulatory debate is focused on safety margins, that is, the smaller the safety margins the bigger the regulatory debate and the greater the amount of proof required to satisfy the regulator. Such experience suggests that caution and prudence guide the development of a regulatory regime for fusion reactors. Fusion has intrinsic safety and environmental advantages over fission, which should alleviate significantly, or even eliminate, the regulatory problems associated with fission. The absence of a criticality concern and the absence of fission products preclude a Chernobyl type accident from occurring in a fusion reactor. Although in a fusion reactor there are large inventories of radioactive products that can be mobilized, the total quantity is orders of magnitude smaller than in fission power reactors. The bulk of the radioactivity in a fusion reactor is either activation products in steel structures, or tritium fuel supplies safely stored in the form of a metal tritide in storage beds. The quantity of tritium that can be mobilized under accident conditions is much less than ten million curies. This compares very favorably with a fission product inventory greater than ten billion curies in a fission power reactor. Furthermore, in a fission reactor, all of the reactivity is contained in a steel vessel that is pressurized to about 150 atmospheres, whereas in a fusion reactor, the inventory of radioactive material is dispersed in different areas of the plant, such that it is improbable that a single event could give rise to the release of the entire inventory to the environment. These intrinsic features give fusion a significant safety and environmental advantage over fission. With such significant intrinsic safety advantages there is noa priori need to make fusion requirements/regulations more demanding and more stringent than fission. To do so could have the effect of making the licensing process more difficult, more costly, and less certain. The result could be a delay in the advent of a safer and more environmentally benign energy system.     

碳化硅基新型包覆燃料颗粒的设计及制备

TRISO型包覆燃料颗粒可将核裂变产生的气体、固体裂变产物束缚在燃料颗粒内部,是高温气冷堆安全性的重要保障。为满足未来超高温气冷堆在更高温度及更高燃耗条件下对燃料元件的要求,需对传统TRISO颗粒进行优化和改进。基于包覆颗粒的破损机制,设计了两种SiC基新型包覆颗粒,一种采用疏松SiC层替代疏松热解炭层,包覆层由内而外依次为疏松SiC层、内致密热解炭层、致密SiC层、外致密热解炭层;另一种为全SiC包覆结构,包覆层由内而外依次为内层疏松SiC层、SiC过渡层、外层致密SiC层。根据结构设计,采用流化床化学气相沉积法实验探索了疏松SiC的形成机制及包覆工艺条件,并利用SEM、XRD等进行材料分析,最终成功实现了两种新型包覆颗粒的大规模制备。更进一步,提出了全SiC基燃料元件的概念,并制备了球形和柱形全SiC基模拟燃料元件。     

Reaction balance and efficiency analysis of a D-D fusion/fission hybrid with satellite D-3He reactors

Selected reactor physics and isotope balance characteristics of a fusion hybrid supported D-³He satellite nuclear energy system are formulated and investigated. The system consists of two types of reactors: a parent D-fueled fusion device and a number of smaller reactors optimized for D-³He fusion. The parent hybrid station breeds the helium-3 for the satellites and also breeds fissile fuel for an existing fission reactor economy. Various hybrid operational regimes are examined in order to determine favorable reactorQ values and effective fusion and fission efficiencies. A number of analytical correlations between power output, plasma energetics, blanket neutronics, breeding capacity, and energy conversion cycles are established and evaluated. Numerical examples of performance parameters such as fission-to-fusion power, overall conversion efficiency, and the ratio of satellite to parent fusion power are presented. The range of reactor efficiencies is elucidated as affected by the internal plasma power balances. As an upper bound based on optimistic injection and direct conversion efficiencies, we find the D-³He satellite system power output attaining at best 1/3 of the parent fusion power.     

包覆燃料颗粒尺寸及其标准偏差对失效概率的影响

三结构同向性型(Tristructural isotropic,TRISO)包覆燃料颗粒是目前高温气冷堆和固态燃料熔盐堆采用的燃料元件。TRISO包覆燃料颗粒破损会导致裂变产物不可接受的释放,由此影响反应堆的安全运行。基于TRISO包覆燃料颗粒压力壳式破损模型,分析了TRISO包覆燃料颗粒核芯和各包覆层的尺寸对失效概率的影响,研究了TRISO包覆燃料颗粒核芯半径、疏松热解碳(Buffer)层厚度和碳化硅(Si C)层厚度的合理设计范围。同时,利用随机抽样统计的方法分析了TRISO包覆燃料颗粒核芯半径分布和各包覆层厚度分布对颗粒失效概率的影响。研究发现,降低Buffer层厚度分布的标准差至16μm可以使TRISO包覆燃料颗粒的失效概率降低一个数量级。     

Environmental and Safety Aspects of Using Tritium in Fusion

Today most current fusion research and development activity is based on the expectation that the D–T reaction will be used for the first generation of fusion reactors. This mixture is the premier candidate for a fusion fuel on account of its outstanding energy gain. Fusion reactors will produce neither the problematic emissions now experienced from fossil-fuel-burning power plants nor the long-lived fission products and transuranic elements resulting from fission reactors. Even though, tritium is a radioactive isotope of hydrogen with 12.32 years half-life that exposes beta radiation. Although its specific activity is relatively weak but because of its gaseous state, it can leak easily from its container and contaminate its surrounding. In order to improving safety and reliability of fusion reactors, research groups jointly investigate radiation hazards resulting from the release of tritium and activation products during normal operations as well as accidental conditions. In this paper, some of the most significant safety and environmental aspects of tritium is briefly discussed.     

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation’s energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.     

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.     

HTGR包覆燃料颗粒碳化硅层细晶化研究

高温气冷堆(HTGR)是能适应未来能源市场的第四代先进核反应堆堆型之一,其固有安全性的第一道保障是使用的TRISO型包覆燃料颗粒。在TRISO型燃料颗粒4层包覆结构中,SiC包覆层是承受包覆燃料颗粒内压和阻挡裂变产物释放的关键层,制备高质量SiC包覆层是核燃料领域中的重大问题和关键技术之一。本文介绍高温气冷堆燃料颗粒的基本结构,详述制备SiC包覆层的流化床-化学气相沉积过程,提出SiC层细晶化这一研究方向,并系统阐述包覆燃料颗粒SiC包覆层细晶化的优势。在细晶化SiC材料制备方法方面,系统分析SiC粉体、陶瓷、薄膜和厚膜材料的研究现状,并结合本实验室前期研究成果提出制备细晶SiC包覆层的可行制备策略。     

The tokamak hybrid reactor

At a time when the potential benefits of various energy options are being seriously evaluated in many countries throughout the world, it is both timely and important to evaluate the practical application of fusion reactors for their economical production of nuclear fissile fuels from fertile fuels. The fusion hybrid reactor represents a concept that could assure the availability of adequate fuel supplies for a proven nuclear technology and have the potential of being an electrical energy source as opposed to an energy consumer as are the present fuel enrichment processes. Westinghouse Fusion Power Systems Department, under Contract No. EG-77-C-02-4544 with the Department of Energy, Office of Fusion Energy, has developed a preliminary conceptual design for an early twenty-first century fusion hybrid reactor called the Commercial Tokamak Hybrid Reactor (CTHR) [1]. This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of ‘client’ Light Water Reactor (LWR) Plants. To the depth this study has been performed, no insurmountable technical problems have been identified. The study has provided a basis for reasonable cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources. A nearer-term concept is also defined using a beam driven fusion driver in lieu of the longer term ignited operating mode.     

The fuel element of the HTR-module, a prerequisite of an inherently safe reactor

To retain fission products after postulated accidents, power reactors usually rely on active safety systems inside the primary circuit, such as e.g. redundant shut down systems and multiple redundant decay heat removal systems. The HTR-Module is employing a different approach which relies entirely on the ability of the coated particle to retain all key radio-nuclides as long as a certain maximum fuel element temperature is not exceeded. Consequently, the reactor is designed such that for any postulated accident this maximum fuel element temperature is not reached even without relying on any active safety systems inside the primary circuit, since the decay heat can be removed to an outside heat sink solely by passive means.The paper discusses the experimental results of fission product release from spherical fuel elements for various temperatures. From the tests as well as from statistical considerations it can be concluded that any hazardous radiation dose to the environment can be excluded if the maximum fuel element temperature in the HTR-Module stays below 1600°C.     

Long time experience with the development of HTR fuel elements in Germany

The development of spherical fuel elements for HTR-designs in Germany is discussed. Special attention is given to the development, production and characterization (incl. kernel and coatings) as well as to the irradiation and post-irradiation examination of the different coated particle systems. It has been demonstrated in various irradiation tests which were supplemented by heating tests that for a modular HTR power plant (with a thermal output of 200 MJ s⁻¹) during the specified normal operation as well as in the case of incidents and even accidents, where the maximum fuel temperature will be below 1620 °C, the fission product release is very low. In this context, it must be mentioned that the present coated particle design has not yet been optimized for the combination of high burn-up and high temperature resistance under accident conditions. The TRISO fuel available is a result from fuel development for large HTR's with steam turbines in a time when the modular concept was not yet been invented although its capabilities inspired the design of modular reactors. Thus, there is still a huge potential for improvement of coated particles especially when plutonium or actinide burning is also taken into account.     

A novel concept of QUADRISO particles. Part II: Utilization for excess reactivity control

In high temperature reactors, burnable absorbers are utilized to manage the excess reactivity at the early stage of the fuel cycle. In this paper QUADRISO particles are proposed to manage the initial excess reactivity of high temperature reactors. The QUADRISO concept synergistically couples the decrease of the burnable poison with the decrease of the fissile materials at the fuel particle level. This mechanism is set up by introducing a burnable poison layer around the fuel kernel in ordinary TRISO particles or by mixing the burnable poison with any of the TRISO coated layers. At the beginning of life, the initial excess reactivity is small because some neutrons are absorbed in the burnable poison and they are prevented from entering the fuel kernel. At the end of life, when the absorber is almost depleted, more neutrons stream into the fuel kernel of QUADRISO particles causing fission reactions. The mechanism has been applied to a prismatic high temperature reactor with europium or erbium burnable absorbers, showing a significant reduction in the initial excess reactivity of the core.     

HTR-PM两根一回路连接管断裂的进气事故分析

进气事故是模块式高温气冷堆关注的超设计基准事故之一,石墨氧化腐蚀反应可能导致反射层结构强度减弱、燃料元件完整性和包容裂变产物能力被破坏,以及产生可燃气体等较严重后果。进气事故的分析研究对进一步掌握高温气冷堆的事故特性以及提高反应堆的安全设计具有重要意义。本文基于200MWe球床模块式高温气冷堆示范工程（HTR-PM）的初步设计,假设与一回路压力边界上、下相连的燃料元件进料管和卸料管同时发生断裂,从而形成烟囱效应并导致空气进入堆芯,利用高温气冷堆专用系统分析软件TINTE对自然循环建立及后续的进气腐蚀过程进行了研究,分析了自然循环流量、堆内石墨腐蚀速率、舱室氧气消耗量、燃料元件温度等关键参数的变化。结果表明,即使考虑腐蚀反应的不均匀性,事故后约60h时才会出现首个燃料包覆颗粒裸露现象,燃料元件最高温度峰值低于1620℃的设计限值,保持完好的燃料包覆颗粒仍具有包容放射性裂变产物的能力。同时,如果在相应的时间内采取措施切断进气源,使石墨腐蚀反应不能继续发展,将不会对反应堆的安全造成严重的影响。