A fast-neutron pulse reactor

In June, 1960, criticality was reached on the fast-neutron pulse reactor constructed at the United Institute of Nuclear Studies. The fast-neutron pulse reactor is designed to work under periodic pulse conditions at a mean power output of about 1 kw. The power pulses are developed from the multiplication of prompt neutrons during an interval in which the reactor is in a supercritical state. The half-width of the power pulses is 36 microseconds, and the pulse repetition frequency may be varied between the limits 8–80 pulses/sec.     

Control of the safe operating period of equipment and pipelines in reactor facilities which are in operation or under construction

The basic assumptions for ensuring safe operation of the components of nuclear facilities, based on controlling service lifetime characteristics, are presented. It is shown up for the Du300 RBMK-1000 pipes, which were made of corrosion-resistant austenitic steel, that this technology can be used effectively in operating power generating units. The complex of measures, developed and validated in the last few years, for monitoring and assessing the technical state of weld seams in pipelines, using the safety concepts “leak before rupture” and “prevention of failure” as well as methods for suppressing the proneness of weld seams to form cracks by the corrosion cracking method under stress, has made possible not only safe operation of Du300 pipelines but it is also a basis for optimizing the volumes and periodicity of operational nondestructive monitoring. __________ Translated from Atomnaya énergiya,Vol. 103, No. 1, pp. 61–65, July, 2007.     

Thermohydraulic studies of sodium-cooled reactor facilities

Thermohydraulic studies of reactor facilities with fast reactors are complex experimentally and computationally. Extensive experimental data are obtained on the velocity and temperature profiles, hydrodynamic resistance and heat emission, initial heat section, and interchannel mixing of the coolant in the fuel assemblies. These are used to develop engineering methods of performing thermohydraulic calculations of fuel assemblies as well as computational compute codes. The particulars of the hydrodynamics and heat transfer in intermediate heat exchangers and steam generators of reactor facilities with fast reactors are studied. This has made it possible to validate their thermohydraulic characteristics.     

Heat transfer during boiling of a liquid metal during emergency cooldown of a fast-neutron reactor

The physics of the processes, the characteristics, and the stability of different regimes, of boiling (nucleate, projectile, disperse-ring), which are observed in experiments investigating the boiling of liquid-metal coolant in a model of a fuel assembly for a fast-neutron reactor in the emergency cooldown regime with low circulation velocity, are analyzed. The experimental setup, the, methods for performing measurements, and the experimental data on the boiling of a liquid metal are described. A mathematical model of the process of boiling of a liquid-metal, coolant in a natural-circulation loop is described, and the results of test calculations for regimes with an increase in heating and with sharp pressure drop are prresented. 7 figures, 12 references. State Science Center of the Russian Federration–A. I. Leipunskii Physics and Power Engineering Institute. Translated from Atomnaya énergiya, Vol. 87, No. 5, pp. 337–342, November, 1999.     

自动数字显示的反应堆核功率测量装置

介绍一种自动数字显示的反应堆核功率测量装置的工作原理,设计特点以及主要技术性能。     

超灵敏小型回旋加速器质谱计性能提高与部件改进

综述在把已建成的小回旋加速器建设成一台加速器质谱计的六年的调试与性能提高过程中对原物理设计和设备部件所进行的改进以及SMCAMS的目前性能水平.     

Fuel storage facilities for 200-MW nuclear heating reactor

This paper reflects the thoughts and work concerning considerations and design of 200-MW nuclear heating reactor (NHR-200) developed in Institute of Nuclear Energy Technology (INET), Tsinghua University, China. Due to the fact that the size of heating reactors is limited to the local demands which are generally smaller that the economic reasonable size as compared to those reactors for electricity production, the design of systems for NHR-200 should be specified in accordance with its design characteristics, and simplified as much as possible for economic aim. The nuclear heating reactor has a low power density in the core and that the annual generation period is only about 180 days. Therefore, the total required number of fuel bundles is rather small. Furthermore, in-vessel spent fuel storage is feasible. All these features raise the potential to simplify the fuel storage system for NHR-200. The fuel storage and inspection facilities are described.     

Comparison of nuclear pulse reactor facilities with reactivity-initiated-accident testing capability

Transient testing of advanced nuclear fuels and other structural materials is pivotal for the research, development, and ultimate demonstration of nuclear energy. Transient testing capabilities exist on a global scale, but these various facilities have different operations and design characteristics. Furthermore, the irradiation experiment vehicles (IEV) used in these transient reactor facilities have varying designs depending on the materials and experiment requirements. The advantages and disadvantages of each facility are presented and discussed by comparing the design and capabilities across nuclear transient reactor facilities (TRF). Further, a discussion of a specific IEV from each TRF shows the operational similarities and differences across TRFs. This comparison shows how some TRF designs are best suited for specific research areas and how others are being re-designed with flexible capabilities in mind. This inventory and comparison of global nuclear transient testing capabilities provides insight into the history, current status and future of nuclear fuel and technology development.