Plasma MHD generator for modelling energy conversion in pulsed thermonuclear reactors

Conclsuions The construction of an experimental model for studying MHD energy conversion from a pulsed thermonuclear reactor is a realistic technical task at the present time. Doing this would permit development of a large scale MHD generator module for the typical parameters of the heated working medium in a pulsed thermonuclear reactor.In principle it is possible to obtain an efficiency of at least about 40% with a linear plasma MHD generator. The efficiency of the whole plant might be increased further by utilization of the thermal energy at the outlet of the MHD channel in traditional methods.When such an MHD generator is built difficulties with the behavior of supersonic plasma streams undergoing strong velocity reduction in a channel and the associated gasdynamic problems can clearly be solved successfully by active modification of the boundary layer and appropriate profiling of the MHD channel. Some complications may arise if a regime with time varying magnetic braking is used. Also important is the problem of the behavior of the plasma stream at large magnetic Reynolds numbers (Re_m1).The basic technological problems are these: materials for the MHD channel, cooling arrangements for the channel (especially the critical cross section of the flow path), and pumping off the boundary layer at the electrodes and preventing lithium condensation on the channel walls. Because of the small magnetic field required, construction of the magnet system will clearly not present substantial technical difficulties associated with its size.The most important physical questions as well as a number of technological questions characteristic of this problem may be investigated on a fairly simple model MHD generator with an output power level of 300–500 MW, a pulse duration of 10–20 msec, and a lithium plasma source.Translated from Atomnaya Énergiya, Vol. 39, No. 6, pp. 387–391, December, 1975.     

Analysis of hexcan failures occurring during the simulation of severe accidents in liquid metal cooled reactors

Under the SIMBATH programme the physical phenomena of transient material movement and relocation during severe LMFBR accidents are investigated out-of-pile. In most of the SIMBATH bundle experiments a failure of the wrapper was observed. From the safety point of view this has implications on the issue of propagation. By openings into the inter-subassembly gaps pressure relief and material release are possible. From the development of failure, based on measurements made during the simulation tests, and from post-experiment investigations three types of failure mode have been identified:

• - Melt-through of the wrapper wall by a jet of hot material from a failing pin. This happened very early during the test. Sodium boiling in the annular bypass prior to failure has not been detected.

• - Melt-through in the simulated fuel region by severe ablation due to local crust instability combined with intense heat input from the flowing melt.

• - Melt-through in the simulated breeding regions close to blockages. This failure mode was always observed together with sodium gross boiling in the annular channel, i.e. reduced cooling of the wrapper wall.

No mechanical failure was detected as a result of the stress concentration in the corners of the hexcan walls. The influence of the internal overpressure is restricted mainly to final break-through after severe ablation and drives the material motions after wrapper failure; it does not control wrapper wall failure in these experiments.     

Code validation and development for MHD analysis of liquid metal flow in Korean TBM

Numerical simulation code such as commercial CFD one, ANSYS-CFX was studied and validated with the previous experimental data for magnetohydrodynamics (MHD) analysis in the liquid metal breeder flow of the Korean Test Blanket Module (TBM) considering the ITER operation conditions. And, a system code for safety analysis was developed with the MARS-FR (Multi-dimensional Analysis Reactor Safety for Fusion Reactor) in this study, whose base code was MARS and it was developed for commercial pressurized water reactor and gas cooled reactor in Korea. First, with the selected benchmarking problem with previous experimental studies and analytic solution, ANSYS-CFX was validated and evaluated. It shows a good agreement with the pressure loss prediction in MHD flow. Second, the system code was developed to predict the MHD effect for the one dimensional flow analysis based on a MHD solution which Miyazaki derived and has semi-implicit scheme to a flow direction. This study shows good agreement between ANSYS-CFX MHD prediction and experimental results as a validation, and the developed system code presents good performance of prediction in MHD pressure loss.     

MHD effects on liquid metal film flow

In this paper a stability analysis is carried out for a liquid metal film flowing over an inclined nonconducting chute with coplanar toroidal magnetic field. A successive expansion technique is employed to investigate the nature and growth rate of the MHD instability caused by the long wavelength perturbations. In the unmagnetized limit, the derived result reduces to the well-known stability criterion for the non-conducting fluid case. It is found that for a sufficiently thick, moderately fast film, the stability criterion can be satisfied. However, if considerably higher flow speed is required (for example, to avoid eruptions of hydrogen bubbles formed during the film's exposure to the charged particle bombardment) then the flow can be MHD-wise unstable. Chutes of very narrow width have to be employed in order to achieve stability, which may not be structurally desirable.     

Evolutionary design of reactor vessel assembly for liquid metal cooled battery

The Battery Omnibus Reactor Integral System (BORIS) is being developed as a multipurpose integral fast reactor at the Seoul National University. This paper focuses on developing design methodology for optimizing geometry of the liquid metal cooled reactor vessel assembly. The key design parameters and constraints are chosen considering technical specifications such as thermal limits and manufacturing difficulties. The evolution strategy is adopted in optimizing the geometry. Two objective functions are selected based upon economic and thermohydraulic reasons. Optimization is carried out in the following steps. First, selected design values are supplied to the momentum integral model code to evaluate steady-state mass flow rate and coolant temperature distribution of the reactor vessel assembly utilizing the thermodynamic boundary condition on heat exchanger calculated by the thermodynamics code. Second, the objective function values are calculated and compared against the previous results. The steps are repeated until an optimum value is obtained. Results of the improved design of the reactor vessel assembly are presented and their characteristics are discussed.     

Fusion energy conversion in magnetically confined plasma reactors

One of the most pressing problems of this century is to solve the energy supply problem and in particular the development of fusion energy technology. Fusion powers the Sun and stars, but on Earth is difficult to achieve in a controlled manner. The International Thermonuclear Experimental Reactor (ITER) is the most technologically advanced machine where net energy from fusion is envisaged to be produced. But this will not be easy, since there are still open issues of plasma confinement, reactor materials, fuel supply, and heat removal. Efficient conversion of fusion energy into the thermal energy in a thermonuclear reactor is, therefore, of great technological relevance and in this paper the energy conversion in magnetically confined plasma reactors is addressed. The chamber wall surrounding the plasma is built from the plasma facing components and from the blanket and divertor modules where the fusion energy is converted into the thermal energy, tritium is produced, and the external components of the chamber are shielded from radiation. The useful materials for building the chamber wall components are low neutron activation steels, refractory metal alloys, and carbon fibre and silicon carbide reinforced composites. The suitable coolants of these components are high pressure helium gas and lithium-based liquid metals and molten salts, where the latter can also serve as tritium breeders. Some of these components will be tested in ITER and eventually may be employed for building demonstration fusion power plants envisaged to become operational during the second half of this century. High performance fusion energy conversion concepts being investigated include: Solid and liquid breeder blankets, separately cooled blankets and tritium breeders, high velocity helium jets for cooling plasma facing components, liquid metals flowing along the solid and through the porous metal walls facing the plasma, liquid metals and molten salts flowing through electrically insulated and non-insulated channels of blankets, and liquid metal heat pipes incorporated into the blankets and divertors for augmenting heat removal and achieving high thermal energy conversion efficiencies. The current fusion-to-thermal energy conversion technologies are, however, in an early stage of development and require reduced-activation, long life operation at high temperatures, resistance to plasma disruptions, and low fusion fuel retention materials, and innovative tritium breeding and heat removal concepts for building simple, reliable, safe, and efficient fusion energy technology.     

Characteristic evaluation of electromagnetic flow couplers using a liquid metal MHD analysis code

It is known that electromagnetic flow couplers have high efficiency. In fact, an experiment using a small-scale annulus model shows that it has the maximum efficiency of 60%. However, another experiment using a middle-scale sector model exhibits that the maximum efficiency is less than 20%. In order to analyze this discrepancy, a computer code for solving magnetohydrodynamic flows is developed and is applied to the evaluation of these experiments. The obtained results show that the efficiency of the small-scale annulus model is high as observed in the experiment and the low efficiency observed in the middle-scale sector model is ascribed to the energy loss due to the eddy current arising in the ends of the magnetic region and the electric contact resistance between the stainless steel duct and the copper bus bar.     

The temperature distribution in liquid metal cooled fuel elements

This article presents the results of experimental and analytical studies of temperature distributions and heat transfer in reactor cores with hexagonal fuel element assemblies. The temperatured is tributions in fuel elements in the central region and at the plane walls are analyzed. Some information i s given on the overheating of fuel element cladding under the spacing fins. Effects related to the redistribution of coolant flow rates over the crosssection of the assembly in the transition from turbulent to laminar flow are discussed.Translated from Atomnaya Énergiya, Vol. 22, No. 5, pp. 372–378, May, 1967.     

钠冷快堆单个燃料组件冷却剂沸腾的数值模拟

在正常功率下快堆单个燃料组件的瞬间完全堵流可能会产生相当严重的后果 ,对其后续事故序列及其潜在的破坏能力进行预测是必要的。对模拟这种现象的SCARABEEBE +1实验在包壳流动之前的阶段进行了数值模拟。程序中采用了两流体、六方程模型来描述沸腾及两相流动 ,应用子通道方法来对基本方程进行离散化 ,以半隐数值方法进行了求解。计算结果与实验观测相吻合 ,这表明该程序可以比较准确地预测单个燃料组件在瞬间完全堵流之后 ,包壳流动之前的行为。     

Status of TRIO_U code for sodium cooled fast reactors

Sodium cooled fast reactors (SFRs) have been developed in France for nearly 50 years with successively Rapsodie, Phenix and Superphenix plants. Nowadays, the so-called Astrid prototype is developed in France in the frame of Generation IV deployment. The Astrid project requires thermal hydraulic inputs to support the design and the safety analysis. This paper deals with some thermal hydraulic concerns in the primary circuit: the subassembly, the core, the hot plenum and the cold plenum. The so-called TRIO_U Computational Fluid Dynamic (CFD) code developed at CEA has been progressively adapted to these Astrid concerns. The paper presents the recent improvements, the present status and the remaining challenges for TRIO_U code on each topic. For the subassembly, refined modelling and sub-channel modelling have been developed in parallel. The validation process based on existing experimental data is in progress. A global core modelling including the inter-wrapper region and the connection to the hot plenum is depicted. The need of experimental validation is pointed out. The core outlet region requires refined Large Eddy Simulation computations to predict temperature fluctuations which can induce thermal fatigue. Validation based on sodium experimental data is briefly presented. Thermal stratification in the plenum is a key point for thermal stress analysis on the structures. Validation process includes the comparison to reactor data. Special developments using a Front Tracking method are carried out to deal with free surface and gas entrainment. A methodology including local and global modelling is developed and the validation process is in progress. For decay heat removal situations and especially in natural convection cases, the whole primary vessel – except at the moment the intermediate heat exchangers and the pumps – is modelled with TRIO_U code. Phenix ultimate tests performed in 2009 will be used for the qualification of these particular situations.     

Development of a two-dimensional model for the thermohydraulic analysis of the hot pool in liquid metal reactors

The SSC-K code has been developed at KAERI based on SSC-L originally developed at BNL to analyze loop-type LMR transients. Because the dynamic response of the primary coolant in a pool-type LMR, particularly the hot pool concept, can be quite different from that in the loop-type LMR, major modifications of SSC-L have been made for the safety analysis of the KALIMER. In particular, it is necessary to predict the hot pool coolant temperature distribution with sufficient accuracy to determine the inlet temperature conditions for the IHXs because the temperature distribution of a hot pool can alter overall system response. In this paper a two-dimensional hot pool model is developed and compared with the experimental data. A preliminary evaluation of unprotected loss of heat sink accidents for the KALIMER design with updated SSC-K code has been performed and analyzed.     

CFD analyses of liquid metal flow in sub-channels for Gen IV reactors

Some of the limitations of Reynolds Averaged Navier Stokes (RANS) based Computational Fluid Dynamics CFD codes in computing the flow and temperature field in a rod-bundle are well known. An in-house validation campaign has indicated that the Baseline Reynolds Stress Model (BSL-RSM) with automatic wall treatment is preferred for RANS analyses of a rod-bundle using the CFX code.As a first step in the present paper, the employed CFX code has been assessed with the analyses of a liquid sodium flow in a rod-bundle as in the TEGENA (TEmperatur- und GEschwindigkeitsverteilungen in Stabbündel mit turbulenter NAtriumströmung) experiment. For this RANS analysis, the full cross-section is modelled to avoid numerical issues associated with symmetric boundary conditions.The influence of pitch-to-diameter ratio (p/d) and rod arrangements on thermal-hydraulics is analyzed by applying the assessed modeling approach. For this purpose, rod-bundles with different p/d are arranged in a square and triangular lattice. The computational sub-channels make use of periodic boundary conditions. RANS computed axial velocity normalized with the friction velocity shows the presence of a logarithmic outer region for both arrangements. Similar behavior was reported based on a Large Eddy Simulation (LES) approach. The analyses reveal that the intensity of secondary flow increases with decreasing p/d for both arrangements. RANS analyzed normal Reynolds stresses normalized with centerline velocity in the smallest gap of rod-bundle reveal their anisotropy. Furthermore, the analyses show that the Nusselt numbers increase with p/d for described flow conditions and for both arrangements. Following observations of flow oscillations in a tight lattice rod-bundle as in Hooper's experiment, as a final step, unsteady RANS simulations for hydraulic analyses using a rod-bundle with small p/d are presented with two commercial CFD codes, namely, CFX and STAR-CCM+. In particular, the analysis of Hooper's hydraulics experiment with a tight lattice rod-bundle having a p/d of 1.1 demonstrates the existence of flow oscillations or instabilities as inferred in the experiment.     

Blanket design studies for maximizing the discharge burnup of liquid metal cooled ATW systems

This paper presents the results of neutronic design studies of lead–bismuth eutectic (LBE) and sodium cooled accelerator transmutation of waste (ATW) blankets. These studies have focused primarily on maximizing the discharge burnup under key thermal-hydraulic and material-related design constraints. Subject to the design constraints on the peak linear power, the maximum coolant velocity, the maximum volume fraction of transuranic (TRU) elements in the dispersion fuel, and the peak fast fluence, design studies have been performed for 840 MW ATW blankets. From these studies, it has been found that the unconstrained discharge burnup for a fixed fuel residence time increases monotonically as the fuel volume fraction and blanket size decrease. The results also show that the discharge burnup is proportional to the peak fast fluence. These indicate that the maximum discharge burnup is primarily determined by imposed design constraints. The maximum discharge burnup achievable under the peak fast fluence limit has been found to be ∼28% for the LBE system, and ∼30% for the sodium system. The optimum fuel volume fraction appears to be ∼0.21 and ∼0.32 for LBE and sodium systems, respectively.     

HYSYS软件应用于熔盐冷却系统瞬态分析的可行性研究

系统软件对熔盐堆冷却剂系统的整体模拟、瞬态分析和安全研究起到至关重要的作用。针对化工领域系统仿真软件HYSYS进行二次开发,植入熔盐物性,修改熔盐换热模型,利用软件已有模型(等效回路加热器模型)尝试分析其对熔盐冷却系统仿真的可行性。为验证修改后的软件的可靠性,对中国科学院上海应用物理研究所钍基核能中心硝酸盐熔盐(KNO3-NaNO2-NaNO3 Molten Salt,HTS)实验装置进行了系统仿真模拟,并与实验结果进行了对比。结果显示,扩展后的软件对熔盐冷却系统的分析研究具有较好的适用性。     

Numerical analysis of liquid metal MHD flows through circular pipes based on a fully developed modeling

Magnetohydrodynamics (MHD) laminar flows through circular pipes are studied in this paper by numerical simulation under the conditions of Hartmann numbers from 18 to 10000. The code is developed based on a fully developed modeling and validated by Samad's analytical solution and Chang's asymptotic results. After the code validation, numerical simulation is extended to high Hartmann number for MHD circular pipe flows with conducting walls, and numerical results such as velocity distribution and MHD pressure gradient are obtained. Typical M-type velocity is observed but there is not such a big velocity jet as that of MHD rectangular duct flows even under the conditions of high Hartmann numbers and big wall conductance ratio. The over speed region in Robert layers becomes smaller when Hartmann numbers increase. When Hartmann number is fixed and wall conductance ratios change, the dimensionless velocity is through one point which is in agreement with Samad's results, the locus of maximum value of velocity jet is same and effects of wall conductance ratio only on the maximum value of velocity jet. In case of Robert walls are treated as insulating and Hartmann walls as conducting for circular pipe MHD flows, there is big velocity jet like as MHD rectangular duct flows of Hunt's case 2.     

Multiple reheat helium Brayton cycles for sodium cooled fast reactors

Sodium cooled fast reactors (SFR) traditionally adopt the steam Rankine cycle for power conversion. The resulting potential for water–sodium reaction remains a continuing concern which at least partly delays the SFR technology commercialization and is a contributor to higher capital cost. Supercritical CO₂ provides an alternative, but is also capable of sustaining energetic chemical reactions with sodium. Recent development of advanced inert-gas Brayton cycles could potentially solve this compatibility issue, increase thermal efficiency, and bring down the capital cost sufficiently to compete directly with light water reactors. In this paper, helium Brayton cycles with multiple reheat and intercooling states are presented for SFRs with reactor outlet temperatures in the range of 510–650 °C. The resulting thermal efficiencies range from 39% to 47%, which is comparable with supercritical recompression CO₂ cycles (SCO2 cycle). A systematic comparison between the multiple reheat helium Brayton cycle and the SCO2 cycle is given, considering compatibility issues, plant site cooling temperature effect on plant efficiency, full plant cost optimization, and other important factors. The study indicates that the multiple reheat helium cycle is the preferred choice over SCO2 cycle for sodium cooled fast reactors.     

Plutonium breeding of light water cooled fast reactors

Light water cooled fast reactor with new fuel assemblies (FA) has been studied for high breeding of fissile plutonium. It achieves fissile plutonium surviving ratio (FPSR) of 1.342 (discharge/loading), 1.013 end and beginning of equilibrium cycle (EOEC/BOEC), and compound system doubling time (CSDT) of 95.9 years at the average coolant density of pressurized water reactor (PWR). It is further improved for reduced moderation boiling water reactor (BWR) (RMWR) coolant density. Fissile plutonium surviving ratio reaches 1.397 (discharge/loading), 1.030 (EOEC/BOEC) and CSDT is 37 years. The present study has shown the possibility of breeding at the PWR coolant density and meeting the growth rate of energy demand of advanced countries at the RMWR and Super FR coolant density for the first time. The new FA consist of closely packed fuel rods. The integrity of welding of fuel rods at the top and bottom ends is maintained as the conventional fuel rods. The coolant to fuel volume fraction is reduced to 0.085, one-sixth of that of RMWR. The volume fraction remains unchanged with the diameter of the fuel rod. The thermal hydraulic design of the cores remains for the future study.     

Computational studies of thermoelectric MHD driven liquid lithium flow in metal trenches

The LiMIT system (Lithium/Metal Infused Trenches) is an innovative plasma-facing component for tokamak divertors, recently proposed at the University of Illinois. Thanks to the coupling of two metals having different Seebeck coefficients, the device is able to generate internal thermoelectric currents as a response to an incoming heat flux from the plasma. One of the two metals is liquid lithium and the second metal is a solid composing the trenches (tungsten, or molybdenum, or stainless steel, etc.). Together with the high toroidal magnetic field, the liquid lithium is propelled by a JxB electrodynamic force inside the solid trenches. In the present work we present a numerical characterization of the device. The diffusion–advection of heat is solved together with the Navier–Stokes equations forced by the JxB electrodynamic force, comprising the thermoelectric contribution. We report parametric plots to show the influence of the toroidal magnetic field and of the plasma heat flux. It is found that the average flow velocity of the liquid lithium peaks at a critical magnetic field, always lower than 1.0 T, and then decreases with an inverse law in the range of tokamak-relevant fields. The flow velocity of the lithium increases with a square-root law versus an increasing heat flux. The heat transfer coefficient of the cooling channels is parametrically investigated, revealing that coefficients higher than >4000 W/m² K are needed for the device in order to withstand heat fluxes of 10 MW/m².     

Analysis of gas–liquid metal two-phase flows using a reactor safety analysis code SIMMER-III

SIMMER-III, a safety analysis code for liquid-metal fast reactors (LMFRs), includes a momentum exchange model based on conventional correlations for ordinary gas–liquid flows, such as an air–water system. From the viewpoint of safety evaluation of core disruptive accidents (CDAs) in LMFRs, we need to confirm that the code can predict the two-phase flow behaviors with high liquid-to-gas density ratios formed during a CDA. In the present study, the momentum exchange model of SIMMER-III was assessed and improved using experimental data of two-phase flows containing liquid metal, on which fundamental information, such as bubble shapes, void fractions and velocity fields, has been lacking.

It was found that the original SIMMER-III can suitably represent high liquid-to-gas density ratio flows including ellipsoidal bubbles as seen in lower gas fluxes. In addition, the employment of Kataoka–Ishii’s correlation has improved the accuracy of SIMMER-III for gas–liquid metal flows with cap-shape bubbles as identified in higher gas fluxes. Moreover, a new procedure, in which an appropriate drag coefficient can be automatically selected according to bubble shape, was developed.

Through this work, the reliability and the precision of SIMMER-III have been much raised with regard to bubbly flows for various liquid-to-gas density ratios.