Fuel cycle for a fusion neutron source

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion–fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium–tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m³Pa/s, and temperature of reactor elements up to 650°C). The deuterium–tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.     

Variants of Regenerated Fissile Materials Usage in Thermal Reactors as the First Stage of Fuel Cycle Closing

At present, 240 000 t of spent nuclear fuel (SF) has been accumulated in the world. Its long-term storage should meet safety conditions and requires noticeable finances, which grow every year. Obviously, this situation cannot exist for a long time; in the end, it will require a final decision. At present, several variants of solution of the problem of SF management are considered. Since most of the operating reactors and those under construction are thermal reactors, it is reasonable to assume that the structure of the nuclear power industry in the near and medium-term future will be unchanged, and it will be necessary to utilize plutonium in thermal reactors. In this study, different strategies of SF management are compared: open fuel cycle with long-term SF storage, closed fuel cycle with MOX fuel usage in thermal reactors and subsequent long-term storage of SF from MOX fuel, and closed fuel cycle in thermal reactors with heterogeneous fuel arrangement. The concept of heterogeneous fuel arrangement is considered in detail. While in the case of traditional fuel it is necessary to reprocess the whole amount of spent fuel, in the case of heterogeneous arrangement, it is possible to separate plutonium and ²³⁸U in different fuel rods. In this case, it is possible to achieve nearly complete burning of fissile isotopes of plutonium in fuel rods loaded with plutonium. These fuel rods with burned plutonium can be buried after cooling without reprocessing. They would contain just several percent of initially loaded plutonium, mainly even isotopes. Fuel rods with ²³⁸U alone should be reprocessed in the usual way.     

铅冷快堆M2LFR-1000堆芯燃料管理方案设计

堆芯燃料管理是反应堆设计中极为重要而且复杂的工作,直接影响着堆芯的经济性。目前国内外对于压水堆等传统热堆已有了较为丰富和成熟的燃料管理计算方法,但对于快堆,由于其中子能谱硬,与传统热堆相比有着不同的控制方式和功率分布,快堆的堆芯燃料管理缺乏系统研究。针对中国科学技术大学自主研发的强迫循环冷却的铅基快堆M2LFR-1000,应用SRAC/COREBN软件包进行堆芯燃耗计算,根据燃耗深度提取核素核子密度,计算伪平衡循环参数进行燃料管理预估,然后进行首循环装料、过渡循环和平衡循环燃料管理方案设计。结果表明:对M2LFR-1000堆芯外区燃料换料组件Pu的富集度进行优化,可以延长换料周期到540 d,提高平均卸料燃耗深度;伪平衡循环结果与平衡循环基本一致,伪平衡循环可以用于燃料管理预估。     

关于压水堆产武器级钚的模拟计算

防止核扩散是国际社会共同努力的目标,其中武器级核材料的防扩散是重中之重.钚是反应堆的副产品,如果不计较经济效益,利用铀为核燃料的反应堆都可以生产武器级钚.本文基于日本Takahama-3压水堆建立了五个模型,并进行中子和燃耗计算,得到两种燃料棒产武器级钚的条件、燃料棒轴向的燃耗分布、组件内燃料棒燃耗的变化区间和全堆芯燃料棒径向燃耗分布.基于上述模型和计算数据给出压水堆堆芯内含有武器级钚的准确位置和UO_2燃料棒中武器级钚的产量.这种低燃耗的乏燃料给国际核不扩散带来了巨大风险,国际社会应该加强对此类乏燃料的监管.     

小型模块化熔盐堆钍利用方案

钍基熔盐堆（Thorium Molten Salt Reactor,TMSR）核能系统先导专项的研究目标是研发第四代裂变反应堆核能系统（即钍基熔盐堆）。为充分利用液态燃料熔盐堆的在线添料与在线燃料处理的优势,同时考虑熔盐堆的快速部署,TMSR先导专项部署了小型模块化熔盐堆。考虑燃料处理技术现状及其可能的发展方向,小型模块化熔盐堆钍利用方案采用"三步走"战略。第一阶段采用在线加料与离线处理,实现钍的成规模利用;第二阶段采用在线加料和在线处理（U）与离线处理（MA）的结合,实现钍的高效利用;第三阶段采用在线加料及在线处理全部重金属,实现钍的自持增殖利用。随着"三步走"战略的逐步实施,钍铀燃料循环模式及后处理性能稳步提高,重金属利用率得到明显改善,同时有效降低了卸料毒性。考虑燃料许可容易度和建堆时间,首先为钍利用方案第一阶段布置了三种可能的启堆燃料,分别为低富集铀、低富集铀加钍和233U加钍。计算结果显示:以低富集铀启堆时,燃料循环性能与水堆相当;以233U启堆时,燃料利用率明显高于水堆,且其放射性毒性比水堆低约2个数量级。The missions of the Thorium Molten Salt Reactor (TMSR) Nuclear Energy System are to research and develop the thorium based molten salt reactors (MSR) belonging to the fourth generation of nuclear fission reactor system. A Small modular Molten Salt Reactor (SmMSR) is deployed to make full use of the advantages of online refueling and online reprocessing and to consider the rapid deployment of MSR. An innovative "three-stage" strategy of thorium utilization based on SmMSR is proposed to take the current condition of fuel reprocessing and its future evolution. The first stage can realize the thorium utilization at a large scale with online refueling and off-line processing. The second stage can obtain efficient thorium utilization with online refueling, online processing of uranium and off-line processing of minor actinides (MAs). The third stage is implemented with self-sustaining or breeding mode with online refueling and online processing of all heavy metals. Along with the development of three stages, the utilization of heavy metals will be obviously improved and the radio-toxicity will be significantly reduced. A SmMSR is designed to achieve the goals of the first stage of thorium utilization. And three kinds of nuclear fuel cycles with different startup fuel types (i.e., low enriched uranium (LEU), thorium mixed with LEU (LEU+Th) and thorium mixed with 233U (233U+Th)) are implemented. The results show that the performance for fuel cycle containing LEU is comparable to the pressurized-water reactor (PWR). Meanwhile, the nuclear utilization for that containing 233U is much higher than PWR, and the radio-toxicity for which is lower by ~2 magnitudes than that for PWR.     

Prospects for and problems of using light-water supercritical-pressure coolant in nuclear reactors in order to increase the efficiency of the nuclear fuel cycle

Trends in the development of the power sector of the Russian and world power industries both at present time and in the near future are analyzed. Trends in the rise of prices for reserves of fossil and nuclear fuels used for electricity production are compared. An analysis of the competitiveness of electricity production at nuclear power plants as compared to the competitiveness of electricity produced at coal-fired and natural-gas-fired thermal power plants is performed. The efficiency of the open nuclear fuel cycle and various versions of the closed nuclear fuel cycle is discussed. The requirements on light-water reactors under the scenario of dynamic development of the nuclear power industry in Russia are determined. Results of analyzing the efficiency of fuel utilization for various versions of vessel-type light-water reactors with supercritical coolant are given. Advantages and problems of reactors with supercritical-pressure water are listed.     

Estimation of the Duration of Cooling Time of the SFA of the VVER-1200 Reactor Depending on the Type of Transport Container

Physics of Atomic Nuclei - As the burnup increases, the requirements of nuclear and radiation safety for spent nuclear fuel (SNF) at the subsequent stages of operation grow. The analysis and...     

Hybrid fusion–fission reactor with a thorium blanket: Its potential in the fuel cycle of nuclear reactors

Discussions are currently going on as to whether it is suitable to employ thorium in the nuclear fuel cycle. This work demonstrates that the ²³¹Pa–²³²U–²³³U–Th composition to be produced in the thorium blanket of a hybrid thermonuclear reactor (HTR) as a fuel for light-water reactors opens up the possibility of achieving high, up to 30% of heavy metals (HM), or even ultrahigh fuel burnup. This is because the above fuel composition is able to stabilize its neutron-multiplying properties in the process of high fuel burnup. In addition, it allows the nuclear fuel cycle (NFC) to be better protected against unauthorized proliferation of fissile materials owing to an unprecedentedly large fraction of ²³²U (several percent!) in the uranium bred from the Th blanket, which will substantially hamper the use of fissile materials in a closed NFC for purposes other than power production.     

Importance functions in calculation of fuel burnup in a VVER

Equations that describe fuel burnup in a VVER are given. Equations for the neutron flux density and the content of fission products are presented in the canonical Cauchy form. Such form of representation of equations lends itself well for their use in solving problems related to optimization of the process of fuel burnup in a nuclear reactor. Also given are equations for the importance functions of neutrons and fission products that correspond to the basic system of equation for phase variables.     

行波堆深燃耗的计算特点

行波堆属于新概念堆型,卸料燃耗深度可达400 GWd/tHM,是现有快堆的3~4倍、压水堆的6~8倍,较高的卸料燃耗深度对堆芯物理分析工具计算正确性提出挑战。基于此,以KYLIN-1程序为基础,从能谱、裂变产物核素重要性、燃耗计算误差累积等方面探究行波堆深燃耗计算特点。对典型行波堆六角形组件分析结果表明:低富集度铀组件寿期初、末能谱差别较大,采用单一权重谱制备的多群截面库用于其燃耗计算时,无限增殖系数偏差较大;为保证行波堆深燃耗计算的正确性,燃耗链应包含重要的70种裂变产物核素;行波堆深燃耗计算时,由于燃耗步增多累积的误差较小,无限增殖系数偏差每燃耗步约为0.001%。     

基于真实临界事故的CAACS程序验证

核临界安全是核工业发展过程中最重要的安全问题之一,其中对临界事故的评价和分析是临界事故后屏蔽设计、应急计划等的基础,因此具有重要的研究意义和工程价值。CAACS是自主开发的溶液系统临界事故分析程序,可计算临界事故的裂变次数、裂变功率、温度随时间变化等。在临界基准实验验证的基础上,利用CAACS对2个真实的临界事故进行分析和计算,并与事故估计值进行对比,结果表明,CAACS的计算结果与事故估计值符合较好,可为后处理厂的工程设计提供临界事故分析的技术手段,为后续的临界瞬态研究打下基础。     

不同类型核燃料对热管冷却反应堆燃耗性能的影响

热管冷却反应堆采用固态反应堆设计理念,具有功率密度高、结构紧凑、固有安全性高等特点,在深空探索、深海勘探、偏远地区等场景中具有广阔的应用前景。核燃料作为热管冷却反应堆的重要组成部分,不同类型核燃料在堆芯燃耗分析时会呈现不同的中子学性能。基于美国爱达荷国家实验室(INL)提出的热管冷却反应堆INL Design A,利用清华大学蒙特卡罗中子输运程序RMC (Reactor Monte Carlo code)建立堆芯物理模型,选取UO₂,(U_0.9Pu_0.1)O₂,U-10Zr,U-8Pu-10Zr,UN,UC这6种核燃料开展燃耗计算,分析了不同核燃料、不同功率水平对热管冷却反应堆堆芯燃耗性能的影响。计算结果表明:在堆芯燃耗深度相同情况下(20.8 GW·d·t?1),装载U-8Pu-10Zr燃料的堆芯所需235U富集度最低(9.8%),具有较好的U-Pu增殖性能。堆芯功率处于5 MW的热管冷却反应堆,燃料中241Pu的存在不仅没起到增大堆芯燃耗深度的作用,反而导致堆芯剩余反应性和堆芯寿期末次锕系核素(MAs)的产量增大,影响反应堆的安全性与经济性。因此,对于装载含有Pu燃料的小功率长寿期热管冷却反应堆,需重点关注241Pu对堆芯燃耗性能的影响。     

Fuel breeding in a hybrid muon catalyzed fusion reactor

We discuss the processes of nuclear fuel burnup and plutonium breeding in the uranium blanket of a hybrid mesocatalytic reactor. The time dependence of the nuclear fuel isotopic concentrations was calculated by the BURNFL code. Using a three-dimensional Monte Carlo MORSE-H code the plutonium and tritium breeding coefficients, the fission rates of uranium and plutonium and a specific power distribution in the blanket were computed. The total fusion energy multiplication factor was obtained as a function of the fuel residence time using results of a detailed calculation of the mesocatalytic channel and estimations of the electronuclear channel.     

On the role of fusion neutron source with thorium blanket in forming the nuclide composition of the nuclear fuel cycle of the Russian Federation

The possible role of available thorium resources of the Russian Federation in utilization of thorium in the closed (U–Pu)-fuel cycle of nuclear power is considered. The efficiency of application of fusion neutron sources with thorium blanket for economical use of available thorium resources is demonstrated. The objective of this study is the search for a solution of such major tasks of nuclear power as reduction of the amount of front-end operations in the nuclear fuel cycle and enhancement of its protection against uncontrolled proliferation of fissile materials with the smallest possible alterations in the fuel cycle. The earlier results are analyzed, new information on the amount of thorium resources of the Russian Federation is used, and additional estimates are made. The following basic results obtained on the basis of the assumption of involving fusion reactors with Th-blanket in future nuclear power for generation of the light uranium fraction ^232+233+234U and ²³¹Pa are formulated. (1) The fuel cycle would shift from fissile ²³⁵U to ²³³U, which is more attractive for thermal power reactors. (2) The light uranium fraction is the most “protected” in the uranium fuel component, and being mixed with regenerated uranium, it would become reduced-enrichment uranium fuel, which would relieve the problem of nonproliferation of the fissile material. (3) The addition of ²³¹Pa into the fuel would stabilize its neutron-multiplying properties, thus making it possible to implement a long fuel residence time and, as a consequence, increase the export potential of the whole nuclear power technology. (4) The available thorium resource in the vicinity of Krasnoufimsk is sufficient for operation of the large-scale nuclear power industry of the Russian Federation with an electric power of 70 GW for more than one quarter of a century. The general conclusion is that involvement of a small number of fusion reactors with Th-blanket in the future nuclear power industry of the Russian Federation would to a large extent solve its problems and increase its export potential.     

美国激光惯性约束聚变能源研究综述

简要地介绍了美国激光惯性约束聚变能源( LIFE ) 的研究现状与发展前景。基于美国国家点火装置( NIF ) 的近期进展,美国利弗莫尔实验室提出了激光惯性约束聚变能源设想,并开始了分解研究。设想用新型二极管泵浦固体激光器产生1.4~2.0 MJ 的激光能量,靶丸聚变增益25~30,打靶频率10~15Hz,实现350~500 MW聚变功率,相当于聚变中子源强1.3×1020 ~1.8×1020 n/s。以此驱动次临界裂变包层,使能量再倍增4~10 倍,实现1 GW电功率的输出。采用创新设计的燃料元件,包层可达到90%以上的燃耗深度,形成一个安全、无碳、燃料资源丰富、核废料少、可持续发展的新型核能源系统。In this paper the present study situation and prospect of the American laser-based Inertial Confinement Fusion Energy ( LIFE ) are briefly introduced. It is based on recent progress of National Inertial Facility ( NIF ) and related research have begun. On the assumption of using laser energy of 1.4 to 2.0 MJ, the target fusion gain G=25~30, the repetition rate 10 to 15 Hz, the fusion power of 350 to 500 MW or neutron source power of 1.3×1020 to 1.8×1020 n/s could be achieved. For a sub-critical fission blanket driven by this fusion neutrons power, energy multiplication M of 4~10 and several GW of thermal power could be obtained. By novel design on fuel pins, burnup more than 90% would be achieved for heavy metals in the blanket. Inertial Confinement Fusion-fission energy is a promising concept, which characterized by inherent safety, richness in nuclear fuel resources, minimization of nuclear waste, non-CO2 emitting ,and it is a sustainable energy source.     

Variants of closing the nuclear fuel cycle

Influence of the nuclear energy structure, the conditions of fuel burnup, and accumulation of new fissile isotopes from the raw isotopes on the main parameters of a closed fuel cycle is considered. The effects of the breeding ratio, the cooling time of the spent fuel in the external fuel cycle, and the separation of the breeding area and the fissile isotope burning area on the parameters of the fuel cycle are analyzed.     

Hybrid fusion reactor for production of nuclear fuel with minimum radioactive contamination of the fuel cycle

The paper presents the results of the system research on the coordinated development of nuclear and fusion power engineering in the current century. Considering the increasing problems of resource procurement, including limited natural uranium resources, it seems reasonable to use fusion reactors as high-power neutron sources for production of nuclear fuel in a blanket. It is shown that the share of fusion sources in this structural configuration of the energy system can be relatively small. A fundamentally important aspect of this solution to the problem of closure of the fuel cycle is that recycling of highly active spent fuel can be abandoned. Radioactivity released during the recycling of the spent fuel from the hybrid reactor blanket is at least two orders of magnitude lower than during the production of the same number of fissile isotopes after the recycling of the spent fuel from a fast reactor.     

核电和物理学

20世纪三四十年代，原子物理学家们发现中子轰击铀原子核的裂变现象并首次实现可控裂变链式反应，把世界带进了原子能时代．半个世纪以来核电已达到全世界电力的16％．目前99％以上都是热中子反应堆核电站。在大规模发展核电的情况下，比如说百GWe,必须加快快中子增殖反应堆的发展和推广，方无核燃料匮乏之虞。     

Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium–Uranium–Plutonium Fuel Cycle

It is shown for a closed thorium–uranium–plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain ~54 kg of fission products, ~0.8 kg of thorium, ~0.10 kg of uranium isotopes, ~0.005 kg of plutonium isotopes, ~0.002 kg of neptunium, and “trace” amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.