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.     

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。     

星座型裂变燃料核反应堆的物理构想

从分析钍在持续中子辐照过程中各代子体含量的演变出发,着重研究有多代子体均达到各自的饱和值时的情况和所具有的特性,提出星座型裂变物质核反应堆的物理构想,并就此堆的特性和应用前景作了简单阐述和讨论。     

Analysis of questions concerning the nonproliferation of fissile materials for low-and medium-capacity nuclear power systems

The principles used in the design of low- and medium-power nuclear power systems to decrease the proliferation risk for nuclear materials as well as the results of a comparative analysis of GT-MGR, KLT-40, ABV, and foreign-produced light-water PWRs from this standpoint are presented. The INPRO methodology and modified KAERI for examining the DUPIC fuel cycle are used for nonproliferation assessment. The results show that at the stage where fuel is used in a reactor the resistance to nonproliferation of nuclear materials for the facilities which are now being designed is close to that determined for PWR and is even higher for GT-MGR with low-enrichment uranium fuel. Translated from Atomnaya énergiya, Vol. 105, No. 3, pp. 123–127, September, 2008.     

'Design to manufacture': journey through criticality considerations for new design fissile material transport packages

Abstract

The development of new fissile material transport packages is an increasingly complex and costly business. The IAEA regulations stipulate performance requirements for the behaviour of the package under both normal and accident conditions. The principal performance drivers are: impact/structural, thermal, shielding and criticality safety. The needs of these disciplines can often compete, with refinements becoming necessary to optimise the design of the package. Maximising the payload while achieving the required levels of safety can be time consuming and expensive. Early communication with the criticality safety analysts can frequently speed up the design process. Numerous variables are significant to criticality safety, (e.g. mass, geometry and spacing of fissile material, presence of moderators, presence of fixed poisons, potential damage to package, etc.). Small changes in the design can result in big changes in the reactivity and this may lead to a very large number of scoping calculations. Increased computing power and improved code abilities can make criticality scoping calculations much more efficient. The development of software to enable a single parameterised input to analyse multiple variations of a basic model provides significant benefits in the production of the criticality analysis and to the overall design process. Criticality analysis of a specific design and of variations in that design can now be performed more quickly and efficiently. Through frequent interaction between members of the design team, a 'fluid' design can be created, and an iterative process can then optimise the package configuration. Retaining the criticality input throughout the design, development, prototype manufacture and testing ensures that the key safety principles are progressed to the manufactured package. In turn, the time and cost of the overall project in producing an acceptable package may be much reduced. The paper presents experience from a criticality viewpoint of the development of new package designs for transport of fissile materials, emphasising the benefits of the utilisation of software/computing developments in this process, with statistics illustrating the efficiencies now achievable.     

核聚变装置辐射监控系统研究

核聚变装置在放电实验期间会产生高能中子和γ射线等辐射,部分辐射会通过生物屏蔽物透射或天空反散射而到达工作场所或者环境。设计并实现了可用于全范围(近、中、远程三种应用条件)中子和光子辐射、计算机集中监控系统(RMS),包括对工作场所进行监测的CAN系统、基于一种新的TRP蓝牙树型拓扑Ad Hoc网络的形成机制和路由协议在装置周围区域内移动状态下进行环境辐射监测的无线网络系统、对远程环境进行监测的移动GPRS无线网络系统。     

Some remarks on the fissile isotopes

Data on the fission cross-section of actinides, obtained from the cross-section libraries, support the Fissile Rule. Some observations on the fissile properties of odd and even isotopes are also presented.     

Probability-based design criteria for nuclear structures

This paper presents a methodology for developing probabilistic design criteria for nuclear structures. The main features of the methodology are calibration, second-moment safety index procedure, and use of available data on load and material variabilities. Examples are included to illustrate these concepts.     

A least-squares fit of thermal data for fissile nuclei

A consistent set of 2200 m s^?1 neutron cross sections, Westcott g-factors and fission-neutron yields have been developed and evaluated for ²³³U, ²³⁵U, ²³⁹Puand ²⁴¹Pu. Auxiliary parameters such as ²³³U, ²³⁴U and ²³⁹Pu half-lives and ²⁵²Cf $\text{v}\text{?}$ values are also presented.All available relevant experimental data on fission, absorption, capture, total and scattering cross sections were revised using the most up-to-date reported standards. The latest Cf $\text{v}\text{?}$ data as well as the corrected $\text{v}\text{?}{}_{\mathrm{<mtext>p</mtext>}}$ (prompt) ratios of the above four nuclei were included in the study. The half-lives of the U and Pu nuclides have been reevaluated and were used to reassess experimental data which depended on the α-counting technique. A least-squares fitting program (lsf) was used to obtain a best overall fit and provide an estimate of the sensitivity of the parameters to the quoted uncertainties in experimental data.The recommended fissile thermal parameters listed in Table 36 now constitute a self-consistent set. Problems encountered in earlier evaluations have been found not to be significant in casting any doubts on the consistency of certain types of input data (such as $\text{v}\text{?}$ measurements and observations made in pile and Maxwellian neutron spectra) with the main body of input data. The least-squares results are presented and comparison has been made with ENDF/B-V and other evaluations.     

Electronuclear breeding of nuclear fuel in thorium targets

Conclusions Electronuclear breeding of fuel in thorium and uranium targets differs mainly in that in the case of thorium the neutron flux and the production of the easily fissioning isotope decrease by approximately one third, the total heat release decreases by more than a factor of 2, and at the same time the temperature gradients at the center of the target increase. As we have already mentioned above, if distance within the target is expressed in the units of g/cm², then the spatial distribution of different quantities is found to be close in the uranium and thorium targets. Joint Institute of Nuclear Research, Dubna. Translated from Atomnaya énergiya, Vol. 76, No. 1, pp. 65–71, January, 1994.     

Pinhole collimator design for nuclear survey system

A conventional knife-edge collimator, which is widely used in gamma camera for medical diagnosis, is not suitable for nuclear imaging system because many scattering radiations near the pinhole aperture happen and blur image. A new pinhole collimator, which shapes a channeled aperture for reducing image degradation induced by the scattering radiations, is introduced and its characteristics are analyzed by Monte Carlo simulation. Resolutions defined as the full-width at half-maximum (FWHM) of point spread function and efficiencies are calculated about several pinhole diameters from 4 to 8 mm and channel heights from 2 to 10 mm. For this calculation, we assumed that ¹³⁷Cs radiation sources with 662 keV mono-energies enter into our designed collimator at the 1 m distance from the detector plane. The efficiencies and resolutions of the channeled collimator are compared with those of the conventional collimator. By comparison results, it is verified that the new collimator takes advantage more than the conventional collimator. The optimum channel height and diameter of the pinhole collimator from simulation results are also proposed and designed. We finally acquired nuclear image mounting this collimator in the nuclear survey system.     

核反应技术在高蛋白小麦选种中的应用

利用13．4MeV的氘束在^14N和^12C内引发的（d,p)反应测量了10个品系小麦种子中的^14N/^12C，由此选出高蛋白质含量的小麦种子。     

γ-ray self-absorption of cylindrical fissile material

The self-absorption of γ-ray emitted from cylindrical fissile materials, such as 235U and 239Pu, does not possess spherical symmetry. The analytical formulae of self-absorption for γ-ray throughout the cylinder have been obtained. The intensity of γ-ray is a function of γ-ray outgoing directions and cylindrical configurations, accordingly one can acquire the information about geometrical configuration of cylindrical fissile materials through multi-location measurements. Further more, the method is given in this article. The result can be applied to the fissile material safeguard, such as nuclear monitoring and verifying.     

Nuclear criticality safety considerations for CASTOR casks with spent fuel of the Rossendorf nuclear devices

The spent fuel of the shut down Rossendorf nuclear devices is to be loaded into storage and transport casks of the type CASTOR-MTR-2. According to the variety of different nuclear devices at the Rossendorf site, the Rossendorf fuel is characterized by a great variety with regard to geometry, material, enrichment, and burn-up. According to the special loading conception, the fuel is embedded in aluminium bodies that the fill the CASTOR. The void fraction within the CASTOR is very small resulting in a small water fraction if water flooding is assumed. The criticality safety is proved by MCNP and OMEGA calculations. These are independent codes that use a completely different data base. The results of both codes agree very well demonstrating the reliability of the calculations. Apart from the proof of criticality safety, some interesting features were found mainly as a result of the very small water fraction.     

Standardized seismic design spectra for nuclear plant equipment

Standardized seismic design spectra were developed for the nuclear plant equipment. Such standardized spectra will reduce unnecessary spectrum computation for the individual plants as well as expedite the selection, purchase and qualification of equipment. Most attractive of all, standard equipment such as electric components can be conservatively qualified just once and then used by all plants located in the same seismic zones. Presented here are idealized design spectrum peak envelopes for different types of structure materials and for both the SSE and OBE conditions. Procedures are also provided for the application of the design spectrum peak envelopes to the seismic qualification of equipment either by dynamic analysis or testing.     

新型反应堆可视化设计方法研究

以核聚变反应堆为研究对象,提出了一种参数化的零部件设计为先导,以可变异的零部件组装逻辑为手段的新型的可视设计方法,所有的设计对象采用实体造型技术,可对系统的设计结果进行三维渲染和动态仿真。设计的对象包括偏滤器、中心柱、包层、纵向场、极向场等,以偏滤器为例,部件组成又包括轨道、支体、外靶板、内靶板、能量沉积板、圆顶和翼。所介绍的设计系统为核聚变装置的灵活设计提供了一种有效的手段,同时也将促进核聚变装