Choices

Author(s):  
Robert Pool

During the 1940s and early 1950s, when atomic energy was new, it was common to hear reactors described as nuclear “furnaces.” They “burned” their nuclear fuel and left behind nuclear “ash.” Technically, of course, none of these terms made sense, since burning is a chemical process and a reactor gets its energy from fission, but journalists liked the terminology because it was easy and quick. One loaded fuel into the reactor, flipped a switch, and things got very hot. If that wasn’t exactly a furnace, it was close enough. And actually, the metaphor was pretty good—up to a point. The basement furnace burns one of several different fuels: natural gas or fuel oil or even, in some ancient models, coal. Nuclear reactors can be built to use plutonium, natural uranium, or uranium that has been enriched to varying degrees. Home furnaces have a “coolant”—the air that is circulated through the furnace and out through the rest of the house, carrying heat away from the fire. Reactors have a coolant, too—the liquid or gas that carries heat away from the reactor core to another part of the plant, where heat energy is transformed into electrical energy. There, however, the metaphor sputters out. In a nuclear reactor, the coolant not only transfers heat to a steam generator or a turbine, but it also keeps the fuel from overheating. The coolant in a furnace does nothing of the sort. And most reactors use a moderator to speed up the fission reaction. The basement burner has nothing similar. But the most importance weakness of the furnace metaphor is that it obscured just how many varieties of reactors were possible—and, consequently, obscured the difficult choice facing the early nuclear industry: Which reactor type should become the basis for commercial nuclear power? The possibilities were practically unlimited. The fuel selection was wide. The coolant could be nearly anything that has good heat-transfer properties: air, carbon dioxide, helium, water, liquid metals, organic liquids, and so on.

2021 ◽  
Vol 30 (5) ◽  
pp. 66-75
Author(s):  
S. A. Titov ◽  
N. M. Barbin ◽  
A. M. Kobelev

Introduction. The article provides a system and statistical analysis of emergency situations associated with fires at nuclear power plants (NPPs) in various countries of the world for the period from 1955 to 2019. The countries, where fires occurred at nuclear power plants, were identified (the USA, Great Britain, Switzerland, the USSR, Germany, Spain, Japan, Russia, India and France). Facilities, exposed to fires, are identified; causes of fires are indicated. The types of reactors where accidents and incidents, accompanied by large fires, have been determined.The analysis of major emergency situations at nuclear power plants accompanied by large fires. During the period from 1955 to 2019, 27 large fires were registered at nuclear power plants in 10 countries. The largest number of major fires was registered in 1984 (three fires), all of them occurred in the USSR. Most frequently, emergency situations occurred at transformers and cable channels — 40 %, nuclear reactor core — 15 %, reactor turbine — 11 %, reactor vessel — 7 %, steam pipeline systems, cooling towers — 7 %. The main causes of fires were technical malfunctions — 33 %, fires caused by the personnel — 30 %, fires due to short circuits — 18 %, due to natural disasters (natural conditions) — 15 % and unknown reasons — 4 %. A greater number of fires were registered at RBMK — 6, VVER — 5, BWR — 3, and PWR — 3 reactors.Conclusions. Having analyzed accidents, involving large fires at nuclear power plants during the period from 1955 to 2019, we come to the conclusion that the largest number of large fires was registered in the USSR. Nonetheless, to ensure safety at all stages of the life cycle of a nuclear power plant, it is necessary to apply such measures that would prevent the occurrence of severe fires and ensure the protection of personnel and the general public from the effects of a radiation accident.


Author(s):  
Xiaomeng Dong ◽  
Zhijian Zhang ◽  
Zhaofei Tian ◽  
Lei Li ◽  
Guangliang Chen

Multi-physics coupling analysis is one of the most important fields among the analysis of nuclear power plant. The basis of multi-physics coupling is the coupling between neutronics and thermal-hydraulic because it plays a decisive role in the computation of reactor power, outlet temperature of the reactor core and pressure of vessel, which determines the economy and security of the nuclear power plant. This paper develops a coupling method which uses OPENFOAM and the REMARK code. OPENFOAM is a 3-dimension CFD open-source code for thermal-hydraulic, and the REMARK code (produced by GSE Systems) is a real-time simulation multi-group core model for neutronics while it solves diffusion equations. Additionally, a coupled computation using these two codes is new and has not been done. The method is tested and verified using data of the QINSHAN Phase II typical nuclear reactor which will have 16 × 121 elements. The coupled code has been modified to adapt unlimited CPUs after parallelization. With the further development and additional testing, this coupling method has the potential to extend to a more large-scale and accurate computation.


2007 ◽  
Vol 22 (1) ◽  
pp. 18-33 ◽  
Author(s):  
Anis Bousbia-Salah

Complex phenomena, as water hammer transients, occurring in nuclear power plants are still not very well investigated by the current best estimate computational tools. Within this frame work, a rapid positive reactivity addition into the core generated by a water hammer transient is considered. The numerical simulation of such phenomena was carried out using the coupled RELAP5/PARCS code. An over all data comparison shows good agreement between the calculated and measured core pressure wave trends. However, the predicted power response during the excursion phase did not correctly match the experimental tendency. Because of this, sensitivity studies have been carried out in order to identify the most influential parameters that govern the dynamics of the power excursion. After investigating the pressure wave amplitude and the void feed back responses, it was found that the disagreement between the calculated and measured data occurs mainly due to the RELAP5 low void condensation rate which seems to be questionable during rapid transients. .


2014 ◽  
Vol 989-994 ◽  
pp. 2097-2100
Author(s):  
Zheng Zhang ◽  
Hai Bo He ◽  
Hao Liang Lu

In order to satisfy the calculation requirements of nuclear power plant operating in different conditions, the integration and combination of reactor core computation modules have been proposed. By writing logical language instructions, and then read by interpreter, the engineering designers can make grammatical analysis, lexical analysis, semantic analysis and information extraction. In Linux system environment, the interpreter can fulfill computational tasks based on the actual operating parameters of nuclear power plant. The comparison results indicate that the calculated results obtained by the interpreter language are correct. Therefore, it also demonstrates that the interpreter language is valid.


Author(s):  
Alberto Sáez-Maderuelo ◽  
María Luisa Ruiz-Lorenzo ◽  
Francisco Javier Perosanz ◽  
Patricie Halodová ◽  
Jan Prochazka ◽  
...  

Abstract Alloy 690, which was designed as a replacement for the Alloy 600, is widely used in the nuclear industry due to its optimum behavior to stress corrosion cracking (SCC) under nuclear reactor operating conditions. Because of this superior resistance, alloy 690 has been proposed as a candidate structural material for the Supercritical Water Reactor (SCWR), which is one of the designs of the next generation of nuclear power plants (Gen IV). In spite of this, striking results were found [1] when alloy 690 was tested without intergranular carbides. These results showed that, contrary to expectations, the crack growth rate is lower in samples without intergranular carbides than in samples with intergranular carbides. Therefore, the role of the carbides in the corrosion behavior of Alloy 690 is not yet well understood. Considering these observations, the aim of this work is to study the effect of intergranular carbides in the oxidation behavior (as a preliminary stage of degenerative processes SCC) of Alloy 690 in supercritical water (SCW) at two temperatures: 400 °C and 500 °C and 25 MPa. Oxide layers of selected specimens were studied by different techniques like Scanning Electron Microscope (SEM) and Auger Electron Spectroscopy (AES).


Author(s):  
R. B. Duffey ◽  
I. Pioro ◽  
X. Zhou ◽  
U. Zirn ◽  
S. Kuran ◽  
...  

One of the six Generation IV nuclear reactor concepts is a SuperCritical Water-cooled nuclear Reactor (SCWR), which is currently under development. The main objectives for developing and utilizing SCWRs are to increase the thermal efficiency of Nuclear Power Plants (NPPs), to decrease electrical energy costs, and possibility for co-generation, including hydrogen generation. Atomic Energy of Canada Limited (AECL) and Research and Development Institute of Power Engineering (RDIPE or NIKIET in Russian abbreviations) are currently developing pressure-tube SCWR concepts. The targeted steam parameters at the reactor outlet are approximately 25 MPa and 625°C. This paper presents a survey on modern SuperCritical (SC) steam turbine technology and a study on potential steam cycles for the SCWR plants. The survey reveals that by the time the Gen IV SCWRs are market-ready, the required steam turbine technology will be well proven. Three potential steam cycles in an SCWR plant are presented: a dual-cycle with steam reheat, a direct cycle with steam reheat, and a direct cycle with a Moisture Separator and Reheater (MSR). System thermal-performance simulations have been performed to determine the overall cycle efficiency of the proposed cycles. The results show that the direct cycle with steam reheat has the highest efficiency. The direct cycle with MSR is an alternative option, which will simplify the reactor design at the penalty of a slightly lower cycle efficiency.


Author(s):  
Marija Miletić ◽  
Rostislav Fukač ◽  
Igor Pioro ◽  
Alexey Dragunov

Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuels supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow’s needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts which will utilize thermal neutron spectrum is a Very High Temperature Reactor (VHTR). This reactor concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behaviour within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Centre Rez, Ltd. The purpose of the High Temperature Helium Loop (HTHL) is to simulate technical and chemical conditions of VHTR’s coolant. The loop is intended to serve an as experimental device for fatigue and creep tests of construction metallic materials for gas-cooled reactors and it should be also employed for research in field of gaseous coolant chemistry. The loop will serve also for tests of nuclear graphite, dosing and Helium purification systems. Because the VHTR is a new reactor concept, major technical uncertainties remain relative to helium-cooled advanced reactor systems. This paper summarizes the concept of the HTHL in the Research Centre Rez Ltd., its design, utilization and future plans for experimental setup.


Author(s):  
J H Large

The decision-making process involving the decommissioning of the British graphite-moderated, gas-cooled Magnox power stations is complex. There are timing, engineering, waste disposal, cost and lost generation capacity factors and the ultimate uptake of radiation dose to consider and, bearing on all of these, the overall decision of when and how to proceed with decommissioning may be heavily weighed by political and public tolerance dimensions. These factors and dimensions are briefly reviewed with reference to the ageing Magnox nuclear power stations, of which Berkeley and Hunterston A are now closed down and undergoing the first stages of decommissioning and Trawsfynydd, although still considered as available capacity, has had both reactors closed down since February 1991 and is awaiting substantiation and acceptance of a revised reactor pressure vessel safety case. Although the other first-generation Magnox power stations at Hinkley Point, Bradwell, Dungeness and Sizewell are operational, it is most doubtful that these stations will be. able to eke out a generating function for much longer. It is concluded that the British nuclear industry has adopted a policy of deferred decommissioning, that is delaying the process of complete dismantlement of the radioactive components and assemblies for at least one hundred years following close-down of the plant. In following this option the nuclear industry has expressed considerable confidence that the decommissioning technology required will he developed with passing time, that acceptable radioactive waste disposal methods and facilities will be available and that the eventual costs of decommissioning will not escalate without restraint.


Author(s):  
Zhe Dong ◽  
Xiaojin Huang ◽  
Liangju Zhang

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.


2013 ◽  
Vol 135 (5) ◽  
Author(s):  
Robert A. Leishear

Hydrogen explosions may occur simultaneously with fluid transients' accidents in nuclear facilities, and a theoretical mechanism to relate fluid transients to hydrogen deflagrations and explosions is presented herein. Hydrogen and oxygen generation due to the radiolysis of water is a recognized hazard in piping systems used in the nuclear industry, where the accumulation of hydrogen and oxygen at high points in the piping system is expected, and explosive conditions may occur. Pipe ruptures in nuclear reactor cooling systems were attributed to hydrogen explosions inside pipelines, i.e., Hamaoka, Nuclear Power Station in Japan, and Brunsbuettel in Germany (Fig. 1Fig. 1Hydrogen explosion damage in nuclear facilities Antaki, et al. [9,10–12] (ASME, Task Group on Impulsively Loaded Vessels, 2009, Bob Nickell)). Prior to these accidents, an ignition source for hydrogen was not clearly demonstrated, but these accidents demonstrated that a mechanism was, in fact, available to initiate combustion and explosion. A new theory to identify an ignition source and explosion cause is presented here, and further research is recommended to fully understand this explosion mechanism. In fact, this explosion mechanism may be pertinent to explosions in major nuclear accidents, and a similar explosion mechanism is also possible in oil pipelines during off-shore drilling.


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