核科学百年讲座第三讲核 能与核武器

核能有两个最具影响的应用：一是它的和平利用——核电(第四讲介绍)，另一是它的军事应用——核武器．文章将介绍核能和核武器基本原理以及核燃料生产方法．首先简要介绍核裂变能与聚变能；然后比较详细地说明原子弹的基本原理，包括中子增殖、自持裂变链式反应条件、临界质量和两种形式原子弹的结构原理等，对于氢弹、中子弹的原理仅作简单介绍；最后，对核燃料的生产：铀同位素分离和钚—239的反应堆生产作了原理性的介绍。     

Thermodynamics and phase diagrams of MoPdRu and related systems

During the irradiation of nuclear fuels up to 10% of the original fissile material (uranium and plutonium) is converted into a wide range of fission products. One particular group of fission-product elements, the transition metals Mo, Tc, Ru, Rh and Pd, are not oxidised at the quite low oxygen potentials existing in the fuels, and are precipitated in the fuel in the form of metallic inclusions. The thermodynamic properties and phase diagram of these alloys are of importance, since the precise composition of these alloys, as a function of radial position can provide important information concerning operating conditions inside a fuel pin. For this reason a thorough reassessment of these systems is under way. This paper reports values of the excess-Gibbs energies of mixing in the MoPd, PdRu and MoRu binary systems, and in the MoPdRu ternary. Calculated sections of the MoPdRu system at selected temperatures are presented.     

加速器驱动的核电站──干净现实的核能源

由一台１６００ＭｅＶ的强流质子加速器来辅助驱动一座熔盐核反应堆,从而组建一座新型的核电站．这座电站中裂变核燃料“燃烧”完全,没有长寿命的重锕系与裂片的核废料输出;并可以直接用天然存在的大量钍和贫铀元素作为核燃料来使用．核电站将２０％电能供给加速器运转,８０％电能并入电网．同时电站还可输出十分稀有的稳定同位素和短寿命医用同位素,作为副产品供应市场．这一干净的核能源就是加速器驱动式核反应堆．简称驱动堆;它没有核废料,比自持式核反应堆安全． A new type of nuclear power station can be built by a moltensalt reactor auxiliary driven by a strong neutron source produced by a intensive proton beam with the energy of 1600 MeV from a powerful accelerator. In the power station the nuclear fuels are completely burnt without some long-lived radioactive wastes both of heaVy actinide and fission products. Furthermore the thorium and sub-critical uranium which are massive existence in nature can be used as an available nuclear fuel in...     

利用3—5MeV γ射线测量核裂变数的可行性研究

提出了利用3—5MeV能区γ射线测量核裂变数的方法, 并且用浓缩铀样品堆照实验验证了该方法的可行性. 结果表明: 该方法可以应用于测量核裂变数, 依据通用方法的结果作为参考标准, 该方法测量不确定度小于16%.     

Measurements of Fission and Radioactive Capture Reaction Rates Inside the Fuel of the Ipen/MB-01

This work presents the measures of the nuclear reaction rates along the radial direction of the fuel pellet by irradiation and posterior gamma spectrometry of a thin slice of fuel pellet of UO₂ at 4.3% enrichment. From its irradiation, the rate of radioactive capture and fission had been measured as a function of the radius of the pellet disk using a Ortec GMX HPGe detector. Lead collimators had been used for this purpose. Simulating the fuel pellet in the pin fuel of the IPEN/MB-01 reactor, a thin UO₂ disk is used, being inserted in the interior of a dismountable fuel rod. This fuel rod is then placed in the central position of the IPEN/MB-01 reactor core and irradiated during 1 h under a neutron flux of 5 ×10⁸ n/cm² s. In gamma spectrometry, 10 collimators with different diameters have been used; consequently, the nuclear reactions of radioactive capture that occurs in atoms of ²³⁸U and the fission that occurs on both ²³⁵U and ²³⁸U are measured in function of 10 different regions (diameter of collimator) of the UO₂ fuel pellet disk. Nuclear fission produces different fission products such as ¹⁴³Ce with a yield fission of 5.9% which decay is monitored in this work. Corrections in geometric efficiency due to introduction of collimators on HPGe detection system were estimated using photon transport of MCNP-4C code. Some calculated values of nuclear reaction rate of radioactive capture and fission along the radial direction of the fuel pellet obtained by Monte Carlo methodology, using the MCNP-4C code, are presented and compared to the experimental data showing very good agreement.     

Contrôler et limiter la dispersion des produits radioactifs des centrales électronucléaires en cas d'accident

We examine the role of physical, chemical and physicochemical differences between the consequences of accidents occurring in high-risk technologies (oil extraction, refinery and transport, big chemistry process, space shuttle flying start, airplane, underground mining, high velocity train transportation, etc.) and the major radioactive major accidents in nuclear power plants (NPP) from the viewpoint of their sequels, confined or not confined, in space and time, and then, their serious impact on human societies, neighboring or not. We identify chemical (as volatile species, noble gases, etc.), physical bodies (as aerosols, etc.) containing radioactive nuclei released in these accidents and seek to control and reduce the causes of their diffusion in the atmosphere, migrations in soils and land, transfer in inland waters and oceans. We propose ways of scientific research leading to solutions with a view to get an improved control of these radioactive substances, and in particular:

• A.reducing as much as possible the irradiation duration by removing the fuel elements from the nuclear core of the NPP reactor and extracting all culprit radioactive nuclei;
• B.designing power reactors using liquid fuels that can be frequently or even continuously removed from the core of the reactor. The disputed radioactive substances can, then, be extracted by chemical processes. The purified nuclear fuel fluid may then be returned to the core;
• C.to separate, in a more radical manner, the circuits of the nuclear fuel and that of the cooling fluid, in order to decrease strongly the radioactive phenomena in the latter, easing its design, operation, maintenance and dismantling;
• D.using a fuel of lower atomic number than uranium or plutonium, as thorium. The use in the present-time fuel cycle of actinides with high atomic numbers Z (number of protons) and A (number of nucleons) leads, by successive neutron captures, to very high Z and A pair–pair nuclei (as americium and curium) subject to spontaneous fissions, then emitting neutrons of fission, even when the nuclear chain reaction is stopped. Neutron emissions are a major hazard if their mass production oversizes the laboratory level. It poses difficult, complicated scientific problems for human radioprotection. A way to improve concepts further than A, B, and C consists in using a nuclear fuel whose Z and A numbers are much smaller than those of uranium, for instance thorium. This is concept D studied here.

A substantial part of these lines of these four research routes has been already studied in other contexts, during the last half-century. What can be used is specified here, with the necessary extensions to the fields of physics, chemistry, and physical chemistry.     

Implementation of focused ion beam (FIB) system in characterization of nuclear fuels and materials

Beginning in 2007, a program was established at the Idaho National Laboratory to update key capabilities enabling microstructural and micro-chemical characterization of highly irradiated and/or radiologically contaminated nuclear fuels and materials at scales that previously had not been achieved for these types of materials. Such materials typically cannot be contact handled and pose unique hazards to instrument operators, facilities, and associated personnel. Over the ensuing years, techniques have been developed and operational experience gained that has enabled significant advancement in the ability to characterize a variety of fuel types including metallic, ceramic, and coated particle fuels, obtaining insights into in-reactor degradation phenomena not achievable by any other means. The following article describes insights gained, challenges encountered, and provides examples of unique results obtained in adapting dual beam FIB technology to nuclear fuels characterization.     

Detection efficiency of fission tracks in apatite related to mica external detectors by the spike irradiation technique

Detection efficiency of fission tracks in non-annealed and annealed Durango apatite crystals related to mica external detectors for fission track dating was estimated experimentally using the spike irradiation technique [Suzuki, 2002a. Abstracts of 21st International Conference on Nuclear Tracks in Solids, p. 22]. The relation of the fission track density between apatite samples and mica external detectors is an important parameter for the determination of fission track ages using an absolute approach especially for crystalline minerals. Spike irradiation of uranium-235-induced fission fragments was carried out using a high uranium content glass, GE1497, fixed firmly on the sample surface. Experimental works were carried out for sections parallel and perpendicular to the crystallographic c-axis of apatite. The detection efficiency of fission tracks in non-annealed apatite related to the mica external detectors was estimated to be 0.90 for the prism face and 0.99 for the basal face. The fission track age of Durango apatite averaged for the prism and basal faces was estimated to be 30.6 (±2.5) Ma, which is concordant with reference values reported previously, using the detection efficiency data with a B₀ value of the CN5 monitor glass and a U-238 spontaneous decay constant value of . The results also indicate that the B₀ age calibration method can be used effectively for the absolute evaluation of fission track ages of apatite.     

次临界能源堆物理设计进展

聚变-裂变混合能源堆包括聚变中子源和次临界能源堆,主要目标是生产电能。回顾了国内外混合堆的发展历史,给出混合能源堆设计的边界条件和约束条件,说明次临界能源堆以铀锆合金为燃料、水为冷却剂的设计思想。利用输运燃耗耦合程序 MCORGS 计算了混合能源的燃耗,给出了中子有效增殖因数、能量放大倍数和氚增殖比等物理量随时间的变化。通过分析能谱和重要核素随燃耗时间的变化,说明混合能源堆与核燃料增殖、核废料嬗变混合堆的不同特点。论述了混合堆的热工设计并进行了安全分析。对于燃耗数值模拟程序,通过多家对算,保证其计算结果的可信性。针对次临界能源堆的特点,利用贫铀球壳建立了贫铀聚乙烯装置和贫铀LiH装置,并且专门设计加工了天然铀装置,开展铀裂变率、造钚率、产氚率等中子学积分实验,验证了数值模拟的可靠性。     

Nuclear fuels: design constraints,irradiation induced changes and end of cycle impact

Nuclear fuel is the material in which the fission reactions occur. As such, it has to support many constraints induced by heat generation, fission product creation, as well as interaction with the coolant. These constraints are taken into account and fulfilled by the design. However, during irradiation, many changes are known to occur. Among them, irradiation damage, OD oxidation and changes in chemical or physical characteristics are to be considered. In addition, intermediate storage adds new constraints that the fuel has to support. With emphasis on the water reactor fuels, by far the most important group, a review will be given of major points, focusing on a few specific areas. To cite this article: C. Lemaignan, C. R. Physique 3 (2002) 763–772.     

Nanomechanical testing of ODS steels irradiated with 1 MeV/amu heavy ions

Development of oxide dispersion strengthened (ODS) steels, as candidates for fuel claddings for Gen IV nuclear reactors, requires a comprehensive study of their behaviour under operating conditions. In this work, 1.2 MeV/amu Kr and Xe irradiation was used to simulate fission fragment impact. New irradiation approaches with respect to a non-homogeneous damage profile under ion irradiation were proposed. Hardness profiles of irradiated ODS steels were obtained by continuous stiffness measurements with subsequent analysis of size effects according to the Nix–Gao model. It was found, that heavy ion irradiation leads to hardness saturation in ODS steels in a damage dose range of 0.1–1 dpa. Observed hardening is about 20% and is not connected with the radiation stability of Y–Ti–O and Y–Al–O oxide particles in ODS steels as it was studied by TEM.     

Activation Characteristics of Fuel Breeding Blanket Module in Fusion Driven Subcritical System

Shortage of energy resources and production of long-lived radioactivity wastes from fission reactors are among the main problems which will be faced in the world in the near future. The conceptual design of a fusion driven subcritical system (FDS) is underway in Institute of Plasma Physics, Chinese Academy of Sciences. There are alternative designs for multi-functional blanket modules of the FDS, such as fuel breeding blanket module (FBB) to produce fuels for fission reactors, tritium breeding blanket module to produce the fuel, i.e. tritium, for fusion reactor and waste transmutation blanket module to try to permanently dispose of long-lived radioactivity wastes from fission reactors, etc. Activation of the fuel breeding blanket of the fusion driven subcritical system (FDSFBB) by D-T fusion neutrons from the plasma and fission neutrons from the hybrid blanket are calculated and analysed under the neutron wall loading 0.5 MW/m2 and neutron fluence 15 MW.yr/m2. The neutron spectrum is calculated with the worldwide-used transport code MCNP/4C and activation calculations are carried out with the well known European inventory code FISPACT/99 with the latest released IAEA Fusion Evaluated Nuclear Data Library FENDL-2.0 and the ENDF/B-V uranium evaluated data. Induced radioactivities, dose rates and afterheats, etc, for different components of the FDS-FBB are compared and analysed.     

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.     

Fission fragment damage in zircon

Abstract

X-ray diffraction, electron microscope and optical studies have been : ade on zircons damaged by fission fragments derived from the action of neutrons on uranium impurities in the zircon lattice. A dosage of about 10¹⁴ fission events · cm^?3 was required to produce diffuse x-ray scattering, observable by conventional photographic methods, and a dosage of about 10¹⁶ fission events · cm^?3 appears to be required to render a zircon amorphous. An optical absorption band, generated apparently by γ-rays, was observed near 270 nm in undoped zircon. The valence state of the uranium impurities was unchanged by irradiation with 10¹⁴ fission events · cm^?3, 2 × 10¹⁸ nvt of fast neutrons, or ～10¹⁰ R of γ-rays. A fission event in zircon was deduced to produce ～10⁴ times more displacement damage than an α-recoil atom. Individual fission tracks were observed directly by electron microscopy after dosage of 10¹⁰-10¹³ fission events - cm^?3, the tracks being similar in appearance to those found by other workers in other materials. With increasing dosage, fission fragment uradiation appears to produce a progressive disordering of the lattice, as was previously deduced for the case of α-recoil irradiation, rather than the formation of new crystallographic phases.     

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.     

Diffusion kinetics of gaseous fission products arising under uranium fuel irradiation

A physical mechanism underlying the diffusion migration of gaseous fission products near the triple joints of grain boundaries during nuclear fuel burnout is suggested and mathematically substantiated. The diffusion kinetics is analyzed with regard to the first invariant of the stress tensor for a given structural defect. The results of mathematical simulation are of interest for analysis of nuclear fuel swelling due to the growth of intergranular pores and formation of channels for fission product migration along the grain boundaries.     

Advantages of Production of New Fissionable Nuclides for the Nuclear Power Industry in Hybrid Fusion-Fission Reactors

A concept of a large-scale nuclear power engineering system equipped with fusion and fission reactors is presented. The reactors have a joint fuel cycle, which imposes the lowest risk of the radiation impact on the environment. The formation of such a system is considered within the framework of the evolution of the current nuclear power industry with the dominance of thermal reactors, gradual transition to the thorium fuel cycle, and integration into the system of the hybrid fusion-fission reactors for breeding nuclear fuel for fission reactors. Such evolution of the nuclear power engineering system will allow preservation of the existing structure with the dominance of thermal reactors, enable the reprocessing of the spent nuclear fuel (SNF) with low burnup, and prevent the dangerous accumulation of minor actinides. The proposed structure of the nuclear power engineering system minimizes the risk of radioactive contamination of the environment and the SNF reprocessing facilities, decreasing it by more than one order of magnitude in comparison with the proposed scheme of closing the uranium–plutonium fuel cycle based on the reprocessing of SNF with high burnup from fast reactors.     

Fission decay properties of ultra neutron-rich uranium isotopes

The fission decay of highly neutron-rich uranium isotopes is investigated which shows interesting new features in the barrier properties and neutron emission characteristics in the fission process. ²³³U and ²³⁵U are the nuclei in the actinide region in the beta stability valley which are thermally fissile and have been mainly used in reactors for power generation. The possibility of occurrence of thermally fissile members in the chain of neutron-rich uranium isotopes is examined here. The neutron number N = 162 or 164 has been predicted to be magic in numerous theoretical studies carried out over the years. The series of uranium isotopes around it with N = 154–172 are identified to be thermally fissile on the basis of the fission barrier and neutron separation energy systematics; a manifestation of the close shell nature of N = 162 (or 164). We consider here the thermal neutron fission of a typical representative ²⁴⁹U nucleus in the highly neutron-rich region. Semiempirical study of fission barrier height and width shows that ²⁵⁰U nucleus is stable against spontaneous fission due to increase in barrier width arising out of excess neutrons. On the basis of the calculation of the probability of fragment mass yields and the microscopic study in relativistic mean field theory, this nucleus is shown to undergo exotic decay mode of thermal neutron fission (multi-fragmentation fission) whereby a number of prompt scission neutrons are expected to be simultaneously released along with the two heavy fission fragments. Such properties will have important implications in stellar evolution involving r-process nucleosynthesis.      

Electron energy-loss spectroscopy of anomalous plutonium behavior in nuclear waste materials

Plutonium-enriched layer has been observed in corroded spent uranium oxide fuel (CSNF). These Pu-enriched regions were examined with analytical transmission electron microscopy combined with electron energy-loss spectroscopy (EELS). The enriched region also contained U, Am, Ru, Zr, but only minor enrichment of rare earth elements. The Pu, possibly as Pu(V) according to EELS measurements, was dispersed within re-precipitated uranium oxide (identified as U3O8) nano-crystals between U(VI) secondary phases and the CSNF surface. The U, Pu, and Am enrichment was observed in the corrosion products with tests on different nuclear fuels. This may have implications for the long-term behavior of CSNF under storage in a geologic waste repository. Furthermore, there may be an increased potential for the generation of Pu-bearing colloids from this type of weathered CSNF.     

In situ Raman monitoring of materials under irradiation: study of uranium dioxide alteration by water radiolysis

In situ Raman scattering studies allow following real‐time evolutions of volume or surface structures under extreme conditions. In nuclear materials sciences, ion irradiation‐induced atomic organization modification and water radiolysis are of a major interest. In order to better understand these phenomena, we have developed an in situ versatile portable Raman spectroscopy system coupled with a cyclotron accelerator, allowing monitoring of a solid/liquid interface under irradiation and thus giving access to effects of radiolysis. The different parts of the system and their improvements are described in details. The system efficiency is highlighted by a comparative study of the time dependence of UO₂ surface modification induced, on one hand by contact with water under irradiation by 5 MeV He²⁺ particles, and on the other hand by pure chemical alteration, through contact with a hydrogen peroxide solution. Copyright © 2012 John Wiley & Sons, Ltd.