Space Nuclear Power Plant Research Articles

Optimizing power of a variable-temperature heat reservoir Brayton cycle for space nuclear power plant

Optimizing power of a variable-temperature heat reservoir Brayton cycle for space nuclear power plant

Full-time domain analysis on operating characteristics of free piston Stirling generator

Free piston Stirling generator (FPSG) has been focused increasingly as energy conversion device for space nuclear power plants due to its long life and maintenance-free characteristics. For the purpose of predicting the operating characteristics and describing the inherent mechanisms of FPSG, experiments have been conducted based on a 1 kW prototype. Pressure fluctuations, pistons’ displacements, phase angle between displacer and power piston, hot and cold end temperatures, etc. were obtained during the experiments. Moreover, a full-time domain numerical model was proposed based on the physical processes experienced by all components within the generator. The comparisons between numerical results and experimental results showed that the relative errors are within 13.1% for the pistons’ amplitudes, 11.7% for the terminal voltage and 3.6% for the operating frequency. The thermodynamic, kinetic and electrical characteristics of the prototype were further analyzed. Some imperceptible phenomena were revealed, including the transient characteristics of heat absorption and rejection, hysteresis loss on the displacer rod and leakage loss at the balancing orifices. A detailed energy distribution scheme was plotted and it can provide guidance for the design and manufacture of subsequent generators of the same type.

ТЕПЛОФИЗИЧЕСКИЕ ИССЛЕДОВАНИЯ: ОТ ПЕРВОГО СТЕНДА К КРУПНОМАСШТАБНОЙ АТОМНОЙ ЭНЕРГЕТИКЕ

The article presents the main research results of thermal physicists of the IPPE from its inception to the present time. Research results in the areas of heat and mass transfer and hydrodynamics of coolants (liquid metals, water), physical chemistry and technology of liquid metal coolants for nuclear power plants for various purposes (nuclear power plants, nuclear submarines, space nuclear power plants), development codes, innovative projects, non-nuclear technologies for the use of liquid metals, heat pipes, analysis and generalization of thermophysical data are considered in the article. As a result of a large complex of experimental and computational studies, the fundamental physicochemical and thermohydraulic regularities of the coolant - impurities - structural materials - protective gas have been studied, scientific foundations have been created for the use of liquid metal coolants in nuclear power. Studies have been carried out to substantiate the technical and economic characteristics of nuclear fuel for operating, under construction and future NPPs of VVER RP, design solutions for passive safety, technical solutions and hydrogen safety devices, heat removal from the reactor through a steam generator and PHRS in case of beyond design basis accidents. As well as design solutions and safety for NPP designs with BN-1200 reactor with sodium coolant, BREST-OD-300 reactor with lead coolant, SVBR-100 reactor with lead-bismuth alloy, MBIR research reactor. The results of these studies made it possible, together with institutes and design organizations, to scientifically substantiate thermal-hydraulic parameters and highly efficient technological processes, develop and practically implement devices and systems that ensure the successful operation of fundamentally new nuclear power plants cooled by water and liquid metals, with original scientific and technical solutions that had no analogue in world practice. R&D works were carried out to substantiate the innovative project VVER with supercritical pressure, the concept of an electro-nuclear subcritical blanket based on the modular principle of constructing an core with liquid-salt melts of fissile materials, studies of thermal hydraulics, mass transfer of high-temperature sodium and the development of a combined coolant purification system to justify the BN-HT reactor with temperature sodium ~900 °C for hydrogen production. The directions of investigations at the present stage are discussed.

The reactor research and test facility

The paper describes an overview of the history of development and the current state of JSC “SSC RF-IPPE” reactor research and test facility designed for assembly, research and full-scale life energy tests of space nuclear power plants with a thermionic reactor. The leading specialists involved in development and operation of this facility are represented. The most significant technological interfaces and upgrade operations carried out in the recent years are discussed. The authors consider the use of an oil-free pumping system as part of this facility during degassing and life testing. Proposed are up-to-date engineering solutions for development of the automated special measurement system designed to record NPP performance, including volt-ampere characteristics together with thermophysical and nuclear physical parameters of a ground prototype of the space nuclear power plant. Key words: reactor research and test facility, thermionic reactor, life energy tests, oil-free pumping system, automated special measurement system, volt-ampere characteristics.

Transformable Elements of the Heat Removal System of Space Nuclear Power Plant

The paper considers the transformable elements of the space nuclear power plant heat removal system located in the hinge joints of deployment system and ensuring transfer of the radiator from starting to working (orbital) position. Welded plate bellows and metal hoses are considered as transformable elements. The results are presented of computational and experimental studies performed on models of hinge assemblies.

Measurements of the reactor neutron power in absolute units

The neutron power of the reactor of the Yenisei space nuclear power plant is measured in absolute units using the modernized method of correlation analysis during the ground-based tests of the Yenisei prototypes. Results of the experiments are given. The desired result is obtained in a series of experiments carried out at the stage of the plant preparation for tests. The acceptability of experimental data is confirmed by the results of measuring the reactor neutron power in absolute units at the nominal level by the thermal balance during the life cycle tests of the ground prototypes.

Parametric analysis of space nuclear power plants in thermodynamic design variables

The purpose of the study is to analyze the influence of efficiency factor and medium-entropy temperatures on the specific mass of a space power plant and the specific area of the radiator cooler with regard for heat losses and mass coefficients of subsystems. Functions of the specific performance sensitivity to variations of design variables are used as the parametric analysis tool. It has been shown that sensitivity functions represent criteria relations that define the optimality and similarity range for space nuclear power plants (SNPP) of different types. To specify the permissible intervals for the variation of design variables, a form has been proposed for recording the specific characteristics with explicit interrelationships between target functions and design variables. The obtained results demonstrate a single-extremum dependence of the considered specific characteristics on generalized design variables. This makes it reasonable to define the optimization problem based on the said criteria of the SNPP technical level.

Параметрический анализ космических ядерных энергетических установок в термодинамических проектных переменных

Параметрический анализ космических ядерных энергетических установок в термодинамических проектных переменных

Measuring flow and pressure of lithium coolant under developmental testing of a high-temperature cooling system of a space nuclear power plant

Sub-megawatt space NPP use lithium as a coolant and niobium alloy as a structural material. In order to refine the lithium-niobium technology of the material and design engineering, lithium-niobium loops were worked out in RSC Energia, and they were tested at a working temperature of lithium equal to 1070–1300 K. In order to measure the lithium flow and pressure, special gauges were developed, which made possible the calibration and checkout of the loops without their dismantling. The paper describes the architecture of the electromagnetic flowmeter and the electromagnetic vibrating-wire pressure transducer (gauge) for lithium coolant in the nuclear power plant cooling systems. The operating principles of these meters are presented. Flowmeters have been developed for channel diameters ranging from 10 to 100 mm, which are capable of measuring lithium flows in the range of 0.1 to 30 L/s with the error of 3% for design calibration and 1% for volume graduation. The temperature error of the pressure transducers does not exceed 0.4% per 100 K; the nonlinearity and hysteresis of the calibration curve do not exceed 0.3 and 0.4%, respectively. The transducer applications are illustrated by the examples of results obtained from tests on the NPP module mockup and heat pipes of a radiation cooler.

Uranium and cesium diffusion in fuel cladding of electrogenerating channel

The results of reactor tests of a carbonitride fuel in a single-crystal cladding from a molybdenum-based alloy can be used in substantiating the operational reliability of fuels in developing a project of a megawatt space nuclear power plant. The results of experimental studies of uranium and cesium penetration into the single-crystal cladding of fuel elements with a carbonitride fuel are interpreted. Those fuel elements passed nuclear power tests in the Ya-82 pilot plant for 8300 h at a temperature of about 1500°C. It is shown that the diffusion coefficients for uranium diffusion into the cladding are virtually coincident with the diffusion coefficients measured earlier for uranium diffusion into polycrystalline molybdenum. It is found that the penetration of uranium into the cladding is likely to occur only in the case of a direct contact between the cladding and fuel. The experimentally observed nonmonotonic uranium-concentration profiles are explained in terms of predominant uranium diffusion along grain boundaries. It is shown that a substantially nonmonotonic behavior observed in our experiment for the uranium-concentration profile may be explained by the presence of a polycrystalline structure of the cladding in the surface region from its inner side. The diffusion coefficient is estimated for the grain-boundary diffusion of uranium. The diffusion coefficients for cesium are estimated on the basis of experimental data obtained in the present study.

Thermodynamic Analysis of a Space Nuclear Power Plant with a Gas-Turbine Energy Conversion System

The results of a thermodynamic analysis of a space nuclear power plant with a gas-turbine energy conversion system are presented. Questions concerning the choice of optimal efficiency and maximum surface area of the radiant cooler are examined. The effect of losses in individual elements of the loop on the main plant performance indices is investigated.

Probabilistic Risk Management Using Risk-Based Safety Goals for the Design of Spacecraft with Onboard Nuclear Reactor Systems

A risk-management framework for space mission launches of nuclear reactors is presented in this paper. The framework is based on a set of risk-based safety goals and relies on decision-theoretic principles that advance system design from concept through operation. Because time-dependent behavior is inherent in space missions, a quasi-dynamic probabilistic risk assessment framework is described. We illustrate a use of the framework with a risk management example.A rationale for, and a trial set of, qualitative safety goals and quantitative design objectives for launching space nuclear power plants are presented. The rationale is based on background risks to the general public, on accident risks to the population in the area of the launch site and on other large-consequence single-event catastrophes. Guidance is also obtained from the safety goals developed by the U.S. Nuclear Regulatory Commission, the U.S. Department of Energy, and the Federal Aviation Administration. The quantitative design objectives developed and presented are also compared to the calculated risks of previous launches with radioisotope thermal-electric generators such as for the Galileo, Ulysses, and Cassini missions.

Space nuclear power plants—I

Space nuclear power plants can be configured in a number of different ways depending on the choice of primary loop transport, conversion techniques and heat addition and rejection systems. The large number of the possible configurations mandates a modular approach. A global approach to analyzing the various configurations has been developed. A generalized software, SNAPS (Space Nuclear Auxiliary Power analysis System), has emerged from the studies thus far. The first part of this project which we are presenting in this paper is this computer program SNAPS designed to simulate space nuclear power plants with different energy conversion techniques, to find optimal characteristics constrained in feasibility limitations, and to design the plant for maximum reliability and flexibility. This program is based on the flowsheeting approach in chemical engineering.

Space nuclear reactors — A post-operational disposal strategy

If 100-kWe and multimegawatt-electric class space nuclear reactors are to play a significant role in humanity's push into cislunar and heliocentric space in the next millennium, the obvious advantages of space nuclear power plants should not be denied to space mission planners due to a failure to develop internationally-acceptable post-operational disposal strategies for spent reactor cores. This is true whether the space reactor has shut down at the end of its normal mission lifetime or in response to an onboard system failure/emergency which causes a premature mission termination. Up until now the great majority of aerospace nuclear safety efforts have concentrated on prelaunch, launch and reactor startup activities. In fact, with the exception of the development of the “nuclear safe orbit” (NSO) concept, little technical attention has yet been given to the post-operational disposal of future space reactors. This paper describes the technical alternatives available for the safe, acceptable disposal of space reactors that could be used in a wide variety of space applications in the 21st Century. Post-operational core radioactivity levels for typical advanced design (hundred kWe-class) space reactors are presented as a function of decay time and contrasted to the spent core radionuclide inventory of the SNAP-10A system, the only nuclear reactor operated in space by the United States. The role of a permanent space station, smart robotic systems, and an operating lunar base in support of spent core disposal strategies is also presented, including use of a selected portion of the lunar surface as an internationally-designated spent reactor core repository.