Effectiveness of hybrid monitoring of radiactive contamination of the atmosphere

Conclusion A method has been proposed for calculation of the errors of estimates of wind velocity and the position of the plumes axis for all points on the axis, and numerical estimates have been given for the effectiveness of hybrid monitoring. The results have been used in plans for automated systems for radiation monitoring of the environment at nuclear power plants.Proektavtomatika State Planning and Design Institute. Energiya Scientific Production Association. Translated from Atomnaya Énergiya, Vol. 71, No. 3, pp. 244–249, September, 1991.     

Improvement of the Technology of Radiometric Enrichment of Uranium Ores

Under the new economic conditions, the problem of decreasing substantially the cost of uranium in the final product obtained by processing mined ores has become acute. A search has shown that the problem can be successfully solved by closely combining hydrometallurgical conversion in plants with inexpensive processes, such as large-batch sorting in radiometric monitoring stations, radiometric piecewise separation, and mound leaching of low-grade ore. Radiometric enrichment produces tailings as well as average and rich concentrates. The rich concentrate will be processed using a separate scheme without a sorption process. A specific example of the combined scheme and the apparatus developed for implementing this scheme are examined. 1 figure, 3 tables, 10 references.     

间距对快堆燃料组件迷宫密封结构性能影响的数值模拟

本文采用计算流体动力学（CFD）方法针对直角梯形齿迷宫密封结构进行了三维数值模拟,对比研究间距对密封性能、流场分布、压力场分布、流速控制的影响,并在最佳间距尺寸下分析了偏心度对密封性能的影响。结果显示：扩大迷宫密封结构间距可有效减少漏流量、提高密封性能,并实现较好的流速控制,还可明显改善近壁面处的流动性能,强化对燃料组件外壁面的冷却;当间距超过3.02 mm时,密封性能提高有限,说明间距超过一定限值时,不能继续通过扩大间距提高迷宫密封结构密封性能,需在结构设计与工程应用中选择合理的间距尺寸;偏心现象导致迷宫密封结构流域内流速、压力分布不均匀,且随偏心度的增大不均匀现象有扩大的趋势,不仅对密封性能产生不利影响,还使得燃料组件振动加剧。     

Computational investigation of the stress-strain state of graphite blocks in light of substantiation of RBMK service-life extension

To increase the service life of RBMK-1000 power-generating units to 45 years, it is necessary to substantiate lifetime of the reactor structures operating under irradiation. Specifically, computational work is needed to substantiate the strength and lifetime characteristics. The characteristic features of each type of reactor, and not only the general behavior of graphite under irradiation, are important for making more accurate predictions. It is desirable to rely on the finite-element model with three-dimensional finite elements, since graphite possesses anisotropic properties in different directions. This article describes a method and program, developed on the basis of the finite-element method, that takes account of the nonuniformity of the neutron field and temperature distributions over the cross-section of a graphite block and the nonuniform anisotropic shrinkage (swelling) and creep of graphite. __________ Translated from Atomnaya énergiya, Vol. 104, No. 3, pp. 157–160, March, 2008.     

Practical aspects of the development of a code of rules for assuring the serviceability of the elements of nuclear power facilities

The 2009 edition of the Federal Law On Technical Regulation, indicating the particulars of technical regulation, makes it possible to implement more fully the provisions of the Federal Law On the Use of Atomic Energy for assuring the required production quality for objects using atomic energy, confirming that the production and associated manufacturing conform to modern standards, including international technologies and procedures. This possibility is realized for equipment and pipelines by means of a domestic code of rules which assure the integrity of the elements of nuclear facilities. A code of rules must preserve the succession of more than 60 years of domestic experience with nuclear power and industry and become a platform for operational adoption of the results of scientific research, and modern industrial technologies.     

加速器驱动洁净能系统中的燃耗行为分析

研究了加速器驱动洁净核能系统（ADS）次临界反应堆内核素的演化。分析结果表明：ADS具有嬗变长寿命核废物的能力。从快堆和热堆的比较可知,ADS的快堆具有输出功率大、长寿命超铀放射性废物的累积水平低、裂变产物对反应堆反应性和能量增益影响小等优点。这些优点在利用U－Pu燃料循环的次临界堆中十分明显。对于利用Th-U燃料循环的次临界堆,热堆和快堆都是可以工作的;而对于U－Pu燃料循环的系统,快堆则是较好的选择。     

Probability of onset of brittle failure of vessels under pressure

Conclusions The method proposed for determining the failure probability of pressurized vessels offers the possibility of quantitative comparison of the safety of different vessels in different operating conditions and estimating their failure probability overall. It may be further developed by replacing determinate characteristics by random characteristics. For example, that q is determinate in Eq. (1) means that the orientation of all the defects in the vessel relative to the largest principal stress is strictly constant. In practice, this means that q must be taken for the defect orientation which is worst from the viewpoint of strength; this makes the method excessively conservative. This may be eliminated if the probability density function of the crack orientation in space is introduced and used to calculate the distribution of values of the random parameter q.Translated from Atomnaya Énergiya, Vol. 64, No. 3, pp. 188–194, March, 1988.     

高温堆非纯氦气中预氧化对3种高温合金的腐蚀行为影响研究

高温气冷堆（简称高温堆）中,由于一回路冷却剂氦气中含有微量（ppm级）不纯杂质,其在高温环境中会对高温堆合金材料造成腐蚀,影响设备的性能。Inconel 617、Hastelloy X、Incoloy 800H是3种高温堆中间换热器及蒸汽发生器设备候选材料。研究表明,镍铬合金在高温下表面生成的富铬氧化层是防止合金在高温下发生严重腐蚀的重要因素。本文对3种合金在高温含杂质氦气中的腐蚀行为进行研究,探究预氧化对3种合金腐蚀行为的影响。并通过称重、扫描电镜、X射线能谱、电子探针显微分析仪以及碳硫分析仪对腐蚀结果进行分析。结果表明,3种合金均出现了不同的氧化和渗碳现象,预氧化对Hastelloy X合金抗腐蚀能力的提升不明显,对Inconel 617合金的抗氧化和渗碳能力有一定提升,对Incoloy 800H合金的抗渗碳腐蚀能力有一定提升。     

Denial of shipment of radioactive material

Abstract

Radioactive sources are used globally in a wide range of beneficial applications in the health care industry, in industrial exploration and development, as well as in basic scientific research and discovery. In fact, some 45% of the world’s medical disposable products are gamma sterilised using cobalt-60 in processing plants located around the world. Other vital applications such as cancer treatment, nuclear medicine, oil exploration and industrial radiography routinely and extensively use radioactive sources. In addition, radionuclides and radiopharmaceuticals are used tens of millions of times every year in the diagnosis and treatment of disease in patients globally. The ability to use such radioactive materials in these sectors is dependent on their safe and secure transport both within and between countries, to ensure that both developing and developed countries have access to this technology in a timely and cost effective manner. However, delay and/or denial of shipment of radioactive materials is having a negative impact on the timely availability of this safe and beneficial technology. What are some of the key reasons for denial of shipment? How serious is this issue and what is its impact on industry and the public? What are international agencies, non-government organisations, regulators and industry doing about it? In 2009 and 2010, International Source Suppliers and Producers Association chaired the International Steering Committee on the Denial of Shipment of Radioactive Material. What are some of the key actions that the International Steering Committee has undertaken? What accomplishments have been achieved and what areas still require international effort among key stakeholders? This paper provides industry perspectives on how the issues causing denial of shipment can be addressed, and how dealing with the root causes of denials will have marked positive impact on the industry and public who rely on them each and every day. Further, this paper provides some examples to show how International Source Suppliers and Producers Association and other industry associations such as Industrial Irradiation Association, Council on Radionuclides and Radiopharmaceuticals and Association of Imaging Producers and Equipment Suppliers have and will continue to be actively involved in addressing this issue.     

Prediction of the lifetime of the elements of the safety and control rods of nuclear reactors

Conclusions Thus the procedure developed can be used to forecast the wear of rubbing elements of the linear electromagnetic mechanisms of the SUZ in nuclear reactors.As the experimental data and operating experience showed, when the SUZ mechanisms are incorrectly assembled, because of misalignments the currents in the control windings must be substantially increased in the reactor setup. This in its turn increases the lateral magnetic attraction forces acting on the armature and increases the wear of the bearing units. The nature of the increase in the lateral forces as a function of the eccentricity is shown in Fig. 3 for different currents i.Thus special attention must be devoted to the quality of the assembly and mounting of the electromagnetic mechanisms of the SUZ in reactor installations. For this reason the procedure presented above for forecasting the lifetime of the elements of SUZ mechanisms is valid if the SUZ mechanisms in nuclear power plants are correctly mounted and operated.Translated from Atomnaya Énergiya, Vol. 62, No. 1, pp. 17–22, January, 1987.     

FELiChEM联锁系统的设计

联锁系统是加速器装置的重要组成部分,用于保护设备和人员的安全。本文基于EPICS设计FELiChEM联锁系统,该系统由硬件联锁系统和软件联锁系统两部分组成,硬件联锁系统又分为机器保护系统(MPS)和人身安全保护系统(PPS)。硬件联锁系统的架构分为IOC层、Profinet IO控制层和Profinet IO设备层。每层均可进行冗余配置,而各层之间相互独立。原型样机的测试显示,硬件联锁系统的响应时间为2.144ms,远好于100ms的设计需求。软件联锁系统的设计采用联锁程序与配置文件分离的方式。测试表明,软件联锁逻辑完全由配置文件确定,具有非常好的灵活性。     

Calculation of emergency situations occurring during the conversion of the research reactor at the Institute of Nuclear Physics of the Academy of Sciences of the Republic of Uzbekistan

Calculations of emergency situations occurring during the conversion of the VVR-SM reactor at the Institute of Nuclear Physics of the Republic of Uzbekistan are presented. Results are presented for the following situations: motion of a compensating rod (two cases), loss of electric power, and obstruction of a fuel assembly. It is shown that safety is maintained for all the emergency situations which were analyzed and the behavior of all zones which were analyzed is actually identical. The computational results will be used for the conversion of VVR-SM to low-enrichment fuel. __________ Translated from Atomnaya énergiya, Vol. 104, No. 6, pp. 339–343, June, 2008.     

Analysis of drop tests of an experimental sample of TUK-84

The calculations of the stress-strain state of the structure of TUK-84 and its drop test for mechanical damage, performed on the SM-é322 KBSM stand, are analyzed. It is shown that the results agree well with one another. Conclusions are drawn concerning the preservation of strength and functionality of TUK-84 after being subjected to mechanical loads under conditions of an accident situation, and recommendations are made concerning the order in which calculations of rolled structures should be performed. A computational model making it possible to analyze the stress-strain state of roll structures with various modifications of TUK-84 without the need of repeating expensive tests for accident conditions during shipment is developed on the basis of the test results. __________ Translated from Atomnaya énergiya, Vol. 100, No. 6, pp. 448–452, June, 2006.     

Status of and possible approaches to the standardization of the safety of elements of the nuclear fuel cycle

Conclusions In summary, the foregoing analysis shows that the systems of standardizing documents for fuel-cycle elements must be developed as quickly as possible. The proposed system must have at the upper levels a hierarchical structure of documents of the regulating agencies and at the lower level it must contain the industry documents. Two principles must be observed: on the one hand, the requirements for the safety of the elements of the nuclear fuel cycle must be satisfied by approaches and principles corresponding to modern approaches to guaranteeing the safety of installations posing a nuclear and radiation hazard and, on the other, the extensive industrial experience accumulated and the corresponding industrial standardizing documentation should be used. An important question in formulating such a system of requirements is: What are the further prospects for the development of the entire nuclear fuel cycle? If the future development of nuclear power in this country is based on a closed fuel cycle, then the entire infrastructure of the elements of the nuclear fuel cycle will have to be modified. The next century, when these enterprises will be operating, will have its own safety requirements, and we must think about this now. If the fuel cycle will be open and several of the presently operating elements of the fuel cycle survive, then we have a different formulation of the problem and different scientific and technical problems. Scientific-Research Center of Gosatomnadzor of the Russian Federation. L. Ya. Karpov Scientific-Research Institute of Physical Chemistry. Institute of Physics and Power Engineering. Gosatomnadzor of the Russian Federation. Translated from Atomnaya énergiya, Vol. 76, No. 4, pp. 264–273, April, 1994.     

Possibility of deceasing the dimension of the problem of optimal design of self-shielded reactors

Conclusions We note that the approach considered in this paper is applicable to the most dangerous accident situations and their combinations for any type of reactor. In addition, reactors of a definite type have their own matrix of unit values of the criteria. For different reactors such matrices can differ not only by accident situations (variants) but also some criteria (safety functionals characterizing a definite accident process). Although for deterministic substantiation of self-shielding of one or another design, it is necessary to examine all accident regimes of reactor operation, at the preliminary-search stage in the general design process attention can be confined to the most significant (dangerous) situations. This will make it possible to decrease substantially the dimension of the problem and increase the efficiency of the investigations. Moscow Engineering-Physics Institute. Translated from Atomnaya énergiya, Vol. 82, No. 3, pp. 178–184, March, 1997.     

Ecological Safety of Underground Leaching of Uranium

The ecological problems arising in open-pit and underground production of solid-mineral deposits are described, the basic advantages of underground leaching, the reasons and sources of contamination of the environment during production of the deposits by underground leaching, and measures for protecting the air, soil, and surface and underground waters, and methods for rehabilitating underground waters are presented. Investigations of natural and artificial neutralization of waste underground-leaching solutions during the interaction of the solutions with technogenically unchanged rocks, performed at one of the uranium deposits, are described. An ecological monitoring scheme is proposed for the Khiagdin deposit.     

外胶凝法制备高温气冷堆UO2核芯的湿法工艺

为制备高温气冷堆用燃料致密UO2核芯,对传统的溶胶-凝胶法进行优化和改进。主要对改进后的外胶凝工艺的湿法部分进行介绍,包括U3O8粉的溶解即欠酸硝酸铀酰(ADUN)溶液的制备、胶液的制备、胶液的分散和胶凝及凝胶球的陈化、洗涤和干燥等,并对湿法过程的机理进行了探讨。采用这一工艺,所得重铀酸铵微球的球形度好、尺寸分布均匀且具有良好空隙结构,经过后续的干法工艺如焙烧、还原和烧结,可制备出合格的高温气冷堆用燃料致密UO2核芯。     

Comparison of three applications of ConTraSPECT

Today, the trend in SPECT attenuation correction is to acquire a transmission data set and reconstruct the corresponding attenuation map, for incorporating into the emission reconstruction. Recently, some techniques have been developed to determine the attenuation map directly from the emission data, by exploiting the fact that in the absence of noise only some specific transmission data can be consistent with given emission data. However it is necessary to include some a priori information so that the consistency conditions can be used in an efficient manner. Here, three types of priors are used: a uniform ellipse, a uniform spline curve, and a uniform spline curve knowing the untruncated part of the transmission sinogram. Once the corresponding attenuation map has been computed, it is used to correct for attenuation in the emission image reconstruction process. The three methods have been tested on both simulations and experimentally acquired data. The results show that the two first methods give equally acceptable results, while the third method provides reconstructions which are closer to reconstructions obtained when the attenuation map is computed from untruncated transmission data     

Heat of formation of UBe13

When a mixture of beryllium power and finely dispersed uranium, prepared by the decomposition of its hydride, was heated, a substance was obtained which consisted in the main of UBe₁₃ and a small quantity of free beryllium. From the difference between the heat of solution of UBe₁₃ and of the corresponding mixture of its components, the heat of formation (–H ₂₉₈ ⁰ )UBe₁₃ was determined. After corrections for impurities were allowed for, the heat of formation was found to be 39.3 ± 3.8 kcal/mole.The authors wish to thank N. T. Chebotarev for the x-ray analysis, T. S. Men'shikov for the metallographic analysis.and V. T. Kharlamov and A. I. Lebedev for the determination of oxygen in the alloy and its components.