Nuclear Electric Power Plants Research Articles

Nuclear Power Plants with Circulating Uranium Hexafluoride Based Fuel. Results of Investigations of Fluid Dynamics and Heat Transfer. Applications, Challenges, and Prospects. An Overview

The propositions of the development of reactor systems using UF6 as the nuclear fuel were put forward in USSR and USA already in the fifties of the last century, while from the beginning of the seventies the UF6-based reactors have become considered as the power sources for spaceborne nuclear power plants (NPP). The application of UF6 circulating in the closed NPP contourmakes it possible to realize the potential advantages of the flow-through design due to the gaseous fuel mobility compared with the existing NPPs with solid cores. At different plant designs, reactor arrangements, and special flow organization in fuel elements the power range from hundreds of kilowatts to tens megawatts can be realized. The areas of application of power plants with circulating UF6 can, in particular, include the spaceborne NPPs of wide power range for electric and plasma rocket engines used in the manned flight to Mars, the reactor-laser with direct pumping of gaseous laser mixtures with nuclear fission products, and the on-ground nuclear electric power plants of new generation with high performance with respect to the fuel cycle and safety. On the basis of an analysis of the results of investigations performed in USSR and USA to the end of the nineties of the last century and presented in this overview it can be concluded that from the standpoint of the physics of working processes and constructional materials resistant in the UF6 environment, there are no insurmountable obstacles for developing the NPPs with circulating UF6. The list of problems, whose solution can favor the further development of this line of research, if it would be implemented, is formulated. In Section 1 the fuel element designs are described and the techniques of flow organization in them are experimentally validated. In Section 2 the computational and experimental support of the programs is presented. It includes the calculations of the fluid dynamics and heat and mass transfer, the thermal and transport properties of uranium hexafluoride, the special properties of uranium hexafluoride as a working body, and the uranium hexafluoride effect on the constructionalmaterials. In Section 3 the projects of spaceborne power plants of closed and open type on a wide power range are reviewed. In Section 4 one of the promising lines in the field of the nuclear energy use, namely, the reactor-laser development, is presented. Section 5 overviews the on-ground plants, where, apart from the electric power plants, of interest are transport power plants, high-temperature engineering systems, in particular, for hydrogen production, the plants for producing high neutron fluxes, and some others. Section 6 presents considerations concerning the realization of rig reactor experiments with uranium hexafluoride circulating in the reactor core, whose criticity is fully ensured by the gaseous UF6. These experiments could be the final stage in corroborating the power source of the new type both under space and on-ground conditions. Section 7 contains an analysis of the available publications on the state of the art of investigations in the United States, which allows one to suppose that in USA a vast program of studies on the use of nuclear reactors with circulating UF6 in the space and on-ground energetics is being systematically performed. In Section 8 some problems which need to be solved in developing the power plants with circulating uranium hexafluoride are listed; they require a vast amount of scientific research. The work on the determination of the design of the power plants with circulating UF6 and hydrodynamic processes occurring in them were conducted in the Keldysh Research Center (before 1976 the Institute of Thermal Processes) under the supervision of A.A. Pavel’ev, who was the author of many ideas in the field of hydrodynamic stability and turbulence realized in experimental setups and computationmethods. In memory of his invaluable contribution into this line of research the authors dedicate to him this overview.

THE ANALYSIS OF STRUCTURAL RELIABILITY OF THE MAIN ELECTRIC CONNECTION CIRCUITS OF NUCLEAR POWER PLANTS

The reliability of the main circuit of electrical connections at a nuclear electric power plant that has two units with a capacity of 1,200 MW each has been determined. Reliability, economical, maneuverable properties of the atomic power plant under study are largely determined by its main circuit, so the choice of the circuit for the design and its status in the process of operation occur to be critical objectives. Main electrical connection circuits in nuclear electric power plants are selected on the basis of the schematic networks of the energy system and the land attached to the plant. The circuit of the connection of a nuclear power plant to the grid in the original normal operating modes at all stages of the construction of such a plant should provide the outcome of the full added capacity of a nuclear power plant and the preservation of its stability in the power system without the influence of the emergency system automatics when any outgoing transmission line is disabled. When selecting the main circuit the individual capacity of the installed units and their number are taken into account as well as the order of development of the plant and power supply system; the voltage on which the power of a plant is delivered; a shortcircuit current for switchgear high voltage and the need for their limitation by circuit means; the most power that can be lost when damage to any switch. A model of reliability of the main circuit of electrical connections is designed to detect all types of accidents that are possible at the coincidence of failures of elements with the repair and operational modes that differs in composition and damageability of the equipment, as well as under conditions of the development of accidents due to failure of operation of devices of relay protection and automation.

Experimental investigation of the hydraulic characteristics of a counter flow wet cooling tower

Thermal and nuclear electric power plants as well as several industrial processes invariably discharge considerable energy to their surrounding by heat transfer. Although water drawn from a nearby river or lake can be employed to carry away this energy, cooling towers offer an excellent alternative particularly in locations where sufficient cooling water cannot be easily obtained from natural sources or where concern for the environment imposes some limits on the temperature at which cooling water can be returned to the surrounding. This paper concerns an experimental investigation of the hydraulic characteristics of a counter flow wet cooling tower. The tower contains a “VGA.” (Vertical Grid Apparatus) type packing which is 0.42 m high and consists of four (04) galvanised sheets having a zigzag form, between which are disposed three (03) metallic vertical grids in parallel with a cross sectional test area of 0.15 m × 0.148 m. The present investigation is focused mainly on the effect of the air and water flow rates on the hydraulic characteristics of the cooling tower, for different inlet water temperatures. The two hydrodynamic operating regimes which were observed during the air/water contact operation within the tower, namely the Pellicular Regime (PR) and the Bubble and Dispersion Regime (BDR) have enabled to distinguish two different states of pressure drop characteristics. The first regime is characterized by low pressure drop values, while in the second regime, the pressure drop values are relatively much higher than those observed in the first one. The dependence between the pressure drop characteristics and the combined heat and mass transport (air–water) through the packing inside the cooling tower is also highlighted. The obtained results indicate that this type of tower possesses relatively good hydraulic characteristics. This leads to the saving of energy.

Analysis of the Ductile-to-Brittle Transition Temperature Shift in a Commercial Power Plant With High Nickel Containing Weld Material

Abstract Plant specific surveillance programs that ideally include all relevant materials and materials combinations that are subjected to neutron irradiation during operation address the degradation due to irradiation of the reactor pressure vessel material for nuclear electric power plants. Plant specific surveillance programs are not unique to the two power plants treated in this study. The current Swedish regulatory system does, however, call for a fairly rigid approach within the surveillance program. In the Swedish case, this means that there is a plant specific predetermined inspection∕test program that has to be followed in order to verify the operability of the power plant and also to verify the operational limits with respect to pressure∕temperature effects on a repetitive basis. The two pressurized water reactor plants Ringhals 3 and 4 have in common that the weld metal used for the butt welds of the reactor pressure vessel is a high nickel type material, above the current limits of the NUREG Reg. Guide 1.99, rev. 2. In the original state, the high nickel content provides excellent fracture toughness in the unirradiated material condition and a low ductile-to-brittle transformation temperature (DBTT). It has, however, been highlighted in several studies that high nickel materials exhibit a very large DBTT shift as a consequence of irradiation, and also that the precipitates that form during the irradiation are not as easily controlled during a heat treatment to remove the irradiation damage as are the copper rich clusters. This paper will present the current state of the art regarding these effects as observed in the weld metal specimens. The paper will present the results from the Charpy V notched and fracture mechanics specimen test encapsulated in the Ringhals Units 3 and 4 surveillance programs. The results from the Ringhals Units 3 and 4 surveillance programs show that there is a need for corrective action to be taken in order to ensure 60 y of operability for the two power plants.

C211 過渡熱伝達分布測定による金属欠損の可視化

In order to realize the reliable inspecting systems, many infrared image sensors are utilized because of their high visual capability not depending on the human visible light condition. Recently, nuclear electric power plant is increasing its usefulness not warm up entire earth system accompanying with human life activity. We are now planning to enhance the infrared inspecting system to carry out the condition based maintenance methodology for nuclear electric power plant. At first, this paper clarifies a relationship between the pixel values constructing infrared image and absolute temperature even though automatic gain control function equipped in infrared CCD camera is activating to get the highest contrast image. Second we demonstrate the absolute temperature visualization, which is the most important inspecting factor in the nuclear power plant, on the bended iron sheet.

U.S. nuclear plant decommissioning funding adequacy — by individual funds, utilities, reactors, and industry-wide — assessed by Monte Carlo and baseline trend methods: 1998, 2000, 2001, and 2004

This paper assesses, over years 1998, 2000, 2001, and 2004 — using varying sets of assumption values — the adequacy of decommissioning funding (year-end) for United States' nuclear electric power plants. I assess each of the 222 individual trust funds and 99 utility owners (as of August 2003), as well as each of the 122 reactors, and industry-wide. Individual trust fund results are the most important because, in general, dollars may not be transferred from one fund to another to strengthen a less-adequately financed fund. Using stochastic (Monte Carlo) analysis (20,000 iterations) to incorporate risk, or uncertainty, in my simulation modeling, I assess funding adequacy for 2004, but with estimated balances and contributions data for 2004. The paper also assesses the baseline (most likely) trends in funding adequacy, using “point” estimates for years 1998, 2000, 2001, and stochastic medians for 2004. I compute two measures of funding adequacy relative to two respective benchmark levels: (1) adequacy of “current year” fund balances; and (2), adequacy of “recent-year” fund contributions. The former is a more “looking-backward” measure relative to a benchmark (i.e., current balance embodies all past year contributions and rates of return on fund assets); the latter, more “looking-forward” relative to a forward benchmark. I use decommissioning fund balances data for 2000 and 1998, as reported by the utilities to the U.S. Nuclear Regulatory Commission (NRC), as well as additional balances, yearly contributions, rates of return, and cost escalation data over 1997–2001 obtained in a 2002 survey of owners by the U.S. Government Accountability Office (GAO). The most fundamental conclusion that one should draw from my simulation modeling results — whether using Monte Carlo or baseline trend analysis — is that these decommissioning adequacy percentages for the 222 funds are extremely variable, although a sizable majority of funds are likely above benchmark. Reduced variability — but still wide — also pertains to the more aggregated results for utilities, reactors, and industry-wide. This variability applies not only at a given point in time, but also across time. Funding adequacies that are substantially below benchmark may be of concern to regulators and other policymakers because of the increased risk of a funding shortfall at scheduled (or premature) permanent plant shutdown. Moreover, when explicitly considering risk — i.e., Monte Carlo analysis — this variability is extremely large. For example, for 2004, while 35% of fund balances — on average — are below benchmark (median values), 82% are below at the lower 5% confidence level but only 8% are below at the upper 5% level. Future outcomes are obviously very uncertain! The public at large, future ratepayers, and/or utilities (i.e., stockholders) may be assessed for funding shortfalls; but, who should bear such responsibility? Have utilities, NRC, and state Public Utility Commissions (PUCs) adequately assessed risk when formulating, and approving, the future funding plans for these 222 individual decommissioning trust funds?

Decrease of suppressor-inducer (CD4+ CD45RA) T lymphocytes and increase of serum immunoglobulin G due to perceived job stress in Japanese nuclear electric power plant workers.

To clarify the effects of perceived job stress on the immune system, a cross-sectional study was conducted in 116 male Japanese workers of a nuclear electric power plant (age, 20 to 39; mean, 31 years). Perceived job stress, i.e., psychological job demand, job control, worksite social support, and job strain, was assessed by means of the Japanese version of the Job Content Questionnaire. The job strain score was calculated as the ratio of the job demand score to the job control score. Blood samples were taken from all workers, and numbers of T and natural killer cell subpopulations, B lymphocytes, total lymphocytes and white blood cells, and serum concentrations of immunoglobulins (IgG, IgM, IgA, IgE and IgD) in their blood were measured. The workers were divided into higher and lower strain groups according to their job strain scores. The number of CD4+ CD45RA+ T lymphocytes in the higher strain group having the job strain score of 0.5 or more (41 workers) was significantly smaller than that in the lower strain group having the score of less than 0.5 (75 workers). In contrast, the serum IgG concentration in the former group was significantly higher than that in the latter group (analysis of covariance with age and smoking as covariates). Also, the numbers of total CD4+ T and total T (CD3+) lymphocytes and of white blood cells in the former group were significantly smaller than those in the latter group. After controlling for age and smoking by the partial correlation coefficient in all 116 workers, the number of CD57+ CD16+ natural killer cells was inversely correlated with job demand and with job strain; the number of CD8+ T lymphocytes was positively correlated with worksite social support; and serum IgG and IgM concentrations were positively correlated with job strain. It is suggested that higher job strain decreases the number of CD4+ CD45RA+ T lymphocytes in male Japanese workers but increases serum IgG concentrations.

In-pipe Mobile Micromachine Using Fluid Power Adaptable to Pipe Diameters

To realize maintenance microrobots for low-diameter pipes of nuclear reactors and electric power plants requires high-power in-pipe mobile micromachines. We proposed a bellows microactuator using fluid power and applied actuators to in-pipe mobile micromachines. Previous papers have described mobile inchworm prototypes 25mm in diameter and travel performance. To miniaturize the in-pipe mobile machine and adapt it to pipe diameters, a support mechanism using a toggle and adaptable to pipe diameters was proposed and support force clarified. We proposed a simple rubber-tube microactuator constrained with a coil spring and described static features. We fabricated a mobile micromachine for pipes 4-5mm in diameter that traveled at 7mm/s both forward and backward.

Inherently safe lead — bismuth — cooled reactors

The design of the BRUS-150 lead-bismuth alloy cooled reactor is outlined. The reactor is a two-loop system with liquid metal coolant in the first loop and a steam-water cycle in the seond loop. One design application described is for a modular, single unit reactor with an electrical output of 150 to 170 MW for multipurpose central nuclear heat and electric power plants. The reactor could also be used in nuclear power plants, special nuclear power plants for utilizing weapons and reactor plutonium, and transmutation of actinides.

Cooperation of the USSR and Czechoslovakia in the construction of a nuclear electric power plant

Cooperation of the USSR and Czechoslovakia in the construction of a nuclear electric power plant

Problems in the Control of a Nuclear Reactor -- Steam Electric Power Plant

The problems of controlling a nuclear reactor-steam electric power plant are associated both with nuclear reactor control and with control of the integrated system, which includes the reactor as a heat source. The problems can be attacked by appropriate use of existing methods for analyzing controlled systems. Existing nuclear power plants demonstrate the feasibility of using these methods. They also indicate that not all the problems are completely solved.