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Investigating the Application of Self-Actuated Passive Shutdown System in a Small Lead-Cooled Reactor
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The application of passively or self-actuated passive safety systems in nuclear reactors allow to simplify the overall plant design, besides improving economics and reliability, which are among the high-level goals set out by the Generation IV International Forum. This thesis focuses on investigating the application of a self-actuated, passive shutdown system for a small, modular lead-cooled fast reactor, and on its implications on the dynamic response to an initiating event. The application of passive shutdown systems for a lead-cooled reactor is not studied extensively, due to the general consensus that lead as a coolant, is too dense to achieve any passive shutdown by gravity. On the contrary, dense liquid lead as a coolant is viewed to be extremely efficient in buoyancy-driven passive shutdown.
Initially neutronic parameters were determined using a combination of Monte Carlo codes, OpenMC and Serpent, by carrying out sensitivity analyses on a critical, hot-state core at middle of life. The reactivity worths of the intended shutdown assemblies and control assemblies were then determined. According to a first-order approximation approach, the passive insertion of shutdown rods was assumed to be influenced by gravity, pressure drag and viscous drag due to flow against the assembly and finally the buoyant force.
Sensitivity analyses were performed for a spectrum of models with varied drag coefficients, in addition to determining the effect of addition of ballast to the assembly and finally to assess the effect of changing coolant flow rate. The time of insertion of the shutdown assembly from its parking position in the core was determined for each of these scenarios. An optimised shutdown foot profile was designed to allow the quickest passive insertion and then implemented in BELLA multi-point dynamics code, in order to perform dynamic analyses of a transient overpower scenario.
This study provides evidence for the viability and reliability of gravity-driven shutdown systems in a heavy liquid metal cooled reactor, and also providing specific data for buoyancy-driven insertion. Further studies could be carried out to investigate the application of such systems in different reactors cooled by, for instance, lead-bismuth or mercury, and also to improve the efficiency of safety systems in sodium cooled reactors.
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... Среди методов снижения этого негативного эффекта наибольшей эффективностью обладает поддержание концентрации кислорода в теплоносителе в заданном диапазоне [14,15]. Этот метод основан на том, что оксиды железа и хрома являются термодинамически более выгодными, чем оксиды свинца и висмута [16]. ...
... ИЗУЧЕНИЕ СКОРОСТИ РАСТВОРЕНИЯ ХРОМИСТОЙ ШПИНЕЛИ В этом разделе исследуется скорость растворения оксидной пленки при различных концентрациях примесей в теплоносителе с использованием разработанной модели. В каждом отдельном случае будет оценено значение концентрации примесей в ядре потока теплоносителя ( ( ) , ( ) , ( ) ), что позволит решить систему уравнений (14) и исследовать полученный результат. ...
... Скорость ядра потока теплоносителя обычно находится в диапазоне от 0.5 м/с до 2.5 м/с. Решение системы (14) в этом случае возможно представить аналитически. Объединяя (14) и (18) мы имеем: ...
В работе сформулирована физико-химическая модель для описания растворения хромистой шпинели в контакте со свинцово-висмутовым теплоносителем в различных услови- ях, характерных для первого охладительного контура реакторов на быстрых нейтронах. Модель основана на идее локального термодинамического равновесия вблизи поверхности хромистой шпи- нели. С ее помощью исследована скорость растворения хромистой шпинели для трех наборов концентраций примесей в теплоносителе. При стандартных условиях работы первого охладитель- ного контура и процессе водородной регенерации скорость растворения оказалась незначитель- ной. В случае отсутствия примесей в теплоносителе скорость растворения достигла 50 микро- метров в год, что значительно превышает критические значения. Это подтверждает необходи- мость поддержания заданного уровня концентрации кислорода в теплоносителе, даже в случае предварительной пассивации стали. Исследовано влияние коэффициентов диффузии на скорость растворения, показано, что изменения в коэффициентах влияют на скорость, но не изменяют принципиального характера результатов.
Lead-cooled fast reactors represent the forefront of Generation IV reactor concepts thanks to the technological development that has taken place over the last 20 years. Lead technology is object of steady attention in many Countries thanks to several development projects and an exponentially increasing industrial interest in recent years. One of the most advanced European projects is focused on the development of ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) and is supported by the FALCON (Fostering ALfred CONstruction) international consortium. The goal is to construct a technology demonstrator which is also prototypic for a lead-cooled-based small modular reactor. To reach this ambitious goal, an important research, development and qualification infrastructure is planned in Romania, leveraging on the know-how cumulated within European and nationally funded endeavors. In particular, the infrastructure will support the scientific program with experimental campaigns dedicated to bridging the technological and scientific gaps, as well as to support the design and licensing of the reactor. This paper presents the configuration of the ALFRED reactor coolant system showing its present state of development and technological innovations. The current plant architecture provides for a thermal power up to 300 MWth and a coolant temperature range from a minimum of 390 °C to a maximum of 520 °C.
The Autonomous Reactivity Control (ARC) is a state-of-the-art, innovative safety system, proposed to be implemented as a self-actuated passive
safety system in Generation IV liquid metal cooled fast reactors. It is
intended to address one of the safety objectives staked out by the Generation IV International Forum; Generation IV nuclear energy systems
operations will excel in safety and reliability. This Master’s thesis studies
the design, implementation and characterisation of an ARC system in a
small lead-cooled fast reactor and intends to demonstrate the contribution to reactor safety during an anticipated transient without SCRAM.
A hot-state model of the core was developed, and the neutronic characteristics were studied using the Serpent2 Monte Carlo code. A model of
the ARC system was developed and implemented in the BELLA multipoint dynamics code, in which analyses of transients were performed. It
was shown that the ARC system provides stringent negative reactivity
feedback during a transient. The steady-state temperatures were reduced
by almost 300 K, compared to an identical transient without the ARC
system. Future investigation and development of the ARC system are of
great interest to the development of reactors cooled by liquid metals. It
can be of particular relevance to developers of sodium reactors currently
facing issues with sodium boiling during transients.
The primary source of greenhouse gas (GHG) emissions are fossil fuels with about 66% share of global electricity generation. Despite the challenges it faces today, nuclear energy is considered an effective technology that can be used in mitigating climate change with specific characteristics that underpin the commitment of some countries to maintain it as a future option. Several studies show the effects of investment minimization policies and the replacement of nuclear power plants with renewables. This implies economic impacts on the price of electricity, which increases the use of fossil fuels resulting in health problems related to air pollution and increasing costs to reduce the carbon emitted in the world. This paper addresses a systematic review of the prospects for nuclear energy investments adopted by countries as a strategic option to mitigate climate change and quantifies a range of carbon dioxide (CO2) emission values that can be avoided using as reference the emission factor of power plants at coal.
We calculate the coefficient of thermal expansion (CTE) in tungsten-rhenium random alloys for Re concentrations between 0% and 50% and for temperatures up to 2400 K by employing the quasiharmonic approximation within the ab initio framework of density functional theory. We treat chemical disorder by the virtual crystal approximation and compute the phonon density of states at two levels of sophistication. While the traditional Debye-Grüneisen (DG) model fails to account for the experimentally observed increase in CTE upon Re addition for concentrations above 10% Re, explicit phonon calculations within density functional perturbation theory lead to an overall good agreement with experiment. Thereby we identify the pronounced phonon softening and anisotropy between transversal and longitudinal modes in W-Re to be responsible for the breakdown of the DG model.
In this paper, the conceptual design of a small lead-cooled nuclear reactor intended to replace diesel-power in off-grid applications is presented. In a vessel of dimensions making it transportable by air, the targeted design performance is to produce 3 MW of electrical power for up to 30 years without reloading of fuel. Consequently, the inner vessel can be sealed, delaying malevolent access to the nuclear fuel and improving security. Alumina forming alloys are applied to ensure long term corrosion protection of fuel cladding tubes, steam generator tubes and primary vessel over the operational temperature regime. Moreover, decay heat can be removed in a completely passive manner by natural convection from the core to the primary coolant and by thermal radiation from the primary vessel to the environment. Finally, the source term is such that relocation of population residing beyond 1 km from the reactor will not be required even in the case of a complete core melt.
The scram function renders the core subcritical in response to plant upsets, which is realized by fast drop of the control rods. In the solid fuel (SF) thorium molten salt reactor system (TMSR), denoted as the TMSR – SF1 project, the drop time of the rod is obviously affected by the driving mechanical drag and the hydraulic resistance of the liquid environment. An inner-house code is developed to analyze the rod drop in the TMSR – SF1. In the code, the total transmission efficiency of the driving mechanism is deduced as a relationship with dropping velocity of the rod. With the experimental and theoretic analysis, the motion process of the rod drop is simulated in this study. And it turns out that the drop time in the molten salt is satisfied with the limiting drop time in the TMSR – SF1. In addition, a hydraulic buffer is designed to alleviate the dropping shock from the scram and the calculation results are presented in the paper as well.
Rhenium diboride (ReB2), containing corrugated layers of covalently bonded boron, is a superhard, metallic compound with a micro-hardness reaching as high as 40.5 GPa (under an applied load of 0.49 N). Tungsten diboride (WB2), which takes a structural hybrid between that of ReB2 and AlB2, where half of the boron layers are planar (as in AlB2) and half are corrugated (as in ReB2), has been shown to not be superhard. Here, we demonstrate that the ReB2-type structure can be maintained for solid solutions of tungsten in rhenium diboride with tungsten content up to a surprisingly large limit of nearly 50 at. %. The lattice parameters for the solid solutions linearly increase along both the a- and c-axes with increasing tungsten content, as evaluated by powder X-ray and neutron diffraction. From micro- and nano-indentation hardness testing, all of the compositions within the range of 0 ‒ 48 at.% W are superhard and the bulk modulus of the 48 at % solid solution is nearly identical to pure ReB2. These results further indicate that ReB2-structured compounds are superhard, as has been predicted from first-principles calculations, and may warrant further studies into additional solid solutions or ternary compounds taking this structure-type.
During operation of nuclear power reactors, reactivity initiated accidents can take place such as a control rod drop. If this occurs, the reactivity increases significantly and leads to an enhancement in power, fuel temperature and damage of reactor eventually. Exact assessment of these accidents depends on the hydrodynamic information. In this research, it is tried to simulate the unsteady flow field around the control rod for a pressurized water reactor power plant. In order to simulate the flow field around the control rod inside the guide tube, averaged Navier–Stokes equations accompanied by the layering dynamic mesh strategy have been used. The information exchange between the two computational stationary and moving grids, the computational grid around the control rod and the grid next to the guide tube, has been taken place through the interface. It was concluded that the time duration of control rod to reach the bottom of the core depends on the leakage. It was also observed that the velocity and acceleration of the control rod would be reduced by decreasing leakage flow rate and in certain leakages, the acceleration of the control rod approaches zero due to equilibrium conditions. During this research, a correlation based on the achieved data was proposed which would provide useful information on the relation between the leakage and the time for control rod to reach the bottom of the core.
This paper describes the current status of the design study and the related research and development for an advanced loop-type fast reactor: the Japan sodium-cooled fast reactor (JSFR). First, the development targets and the major design requirements are established. Then, a cost-down approach in which developing innovative technologies is key to being competitive with another future energy source is discussed.Here, the development status of several innovative technologies such as a two-loop cooling system, reliable reactor system, simplified fuel-handling system, passive reactor shutdown system, mitigation measure against a core disruptive accident, and minor actinide-bearing oxide frei core is described. Last, a review of JSFR development and the demonstration plan for the innovative technologies are discussed.
Nuclear Safety is a comprehensive reference for heath and safety and engineering professionals in the nuclear industry and related bodies, including regulatory authorities, environmental and energy scientists, students, researchers and consultants. It provides the methods and data needed to evaluate and manage the safety of nuclear facilities and related processes using risk-based safety analysis, and provides readers with the techniques to assess the consequences of radioactive releases. The book covers relevant international and regional safety criteria (US, IAEA, EUR, PUN, URD, INI). The contents deal with each of the critical components of a nuclear plant, and provide an analysis of the risks arising from a variety of sources, including earthquakes, tornadoes, external impact and human factors. It also deals with the safety of underground nuclear testing and the handling of radioactive waste. . Covers all plant components and potential sources of risk including human, technical and natural factors . Brings together information on nuclear safety for which the reader would previously have to consult many different and expensive sources . Provides international design and safety criteria and an overview of regulatory regimes . Includes case studies and analysis of major accidents with data and calculations on accompanying website Gianni Petrangeli has 40 years of experience in the nuclear industry. He was Director of the Departmental Area for Integrated Systems and New Technologies of ANPA, the Italian Regulatory body responsible for nuclear and industrial safety. He is a consultant to the United Nations Atomic Energy Agency and a researcher for nuclear safety for the European Commission. He lectures on Nuclear and Industrial Safety at the University of Pisa. * Covers all plant components and potential sources of risk including human, technical and natural factors * Brings together information on nuclear safety for which the reader would previously have to consult many different and expensive sources * Provides international design and safety criteria and an overview of regulatory regimes * Includes case studies and analysis of major accidents with data and calculations on accompanying website.