CFD analysis of the HYPER spallation target

KAERI (Korea Atomic Energy Research Institute) is developing an accelerator driven system (ADS) named HYPER (HYbrid Power Extraction Reactor) for a transmutation of long-lived nuclear wastes. One of the challenging tasks for the HYPER system is to design a large spallation target with a beam power of 15–25 MW. The paper focuses on a thermal–hydraulic analysis of the active part of the HYPER target. Computational fluid dynamics (CFD) analysis was performed by using a commercial code CFX 5.7.1. Several advanced turbulence models with different grid structures were applied. The CFX results reveal a significant impact of the turbulence model on the window temperature. Particularly, the k–ε model predicts the lowest window temperature among the five investigated turbulence models.     

A preliminary design study for the HYPER system

In order to transmute the long-lived radioactive nuclides such as transuranics (TRU), Tc-99, and I-129 in LWR spent fuel, a preliminary conceptual design study has been performed for an accelerator driven subcritical reactor system, called HYPER (HYbrid Power Extraction Reactor). The core has a hybrid neutron energy spectrum which includes fast and thermal neutrons for the transmutation of TRU and fission products, respectively. TRU are loaded into the HYPER core in a TRU–Zr metal form because a metal type fuel has very good compatibility with the pyro-chemical process which retains the self-protection of transuranics at all times. On the other hand, Tc-99 and I-129 are loaded as pure technetium metal and sodium iodide, respectively. Pb–Bi is chosen as a primary coolant because Pb–Bi can provide a good spallation target and produce a very hard neutron energy spectrum. As results, the HYPER system does not need any independent spallation target system. 9Cr–2WVTa is used as a window material because this advanced ferritic/martensitic steel is known to have a good performance in the highly corrosive and radiative environment. The support ratios of the HYPER system are about 4–5 for TRU, Tc-99, and I-129. Therefore, a radiologically clean nuclear power, i.e. zero net production of TRU, Tc-99 and I-129 can be achieved by combining 4–5 LWRs with one HYPER system. In addition, the HYPER system, having good proliferation resistance and high nuclear waste transmutation capability, is believed to provide a breakthrough to the spent fuel problems the nuclear industry is facing with.     

HYPER (Hybrid Power Extraction Reactor): A system for clean nuclear energy

The Korea Atomic Energy Research Institute (KAERI) has been performing accelerator driven system related research and development (RID) called HYPER (HYbrid Power Extraction Reactor) for the transmutation of nuclear waste and energy production through the transmutation process. HYPER program is within the frame work of the national mid and long-term nuclear research plan. KAERI is aiming to develop the elemental technologies for the subcritical transmutation system by the year of 2001 and build a small bench scale test facility (5 MW) by the year of 2006. Some major features of HYPER have been developed and employed. On-power fueling concepts are employed to keep system power constant with a minimum variation of accelerator power. A hollow cylinder-type metal fuel is designed for the on-line refueling concept. Lead–bismuth (Pb–Bi) is adopted as a coolant and spallation target material. 1 GeV 16 mA proton beam is designed to be provided for HYPER. HYPER is to transmute about 380 kg of TRU a year and produce 1000 MW of power. The support ratio of HYPER for LWR units producing the same power is believed to be 56.     

Design and analysis of LMFBR structures

In view of the imperatives concerning the LMFBR structure behaviour, the choice of operating conditions is quite limited. In this sense, some considerations regarding materials and manufacturing, stress analysis and design by test will be given. Much has been realized, but efforts are still necessary in order to attain the final goal.     

非能动安注箱的设计与分析

介绍了非能动安注箱的设计与实验,并用CATHENA程序分析其特性:注入流量的峰值,高注入流量的持续时间,最低注入流量等。计算结果表明非能动安注箱设计满足主要的性能要求,CATHENA程序计算结果与实验数据基本一致,可用于概念设计与事故分析。     

Design and performance analysis of ITER ex-vessel magnetic diagnostics

Accurate magnetic diagnostics are essential to perform reliable operation of any tokamak. The ITER magnetic diagnostics include a wide variety of sensors located on the inner and outer surfaces of the vacuum vessel, in the divertor cassettes and in the casing of the toroidal field coils. As the measurement accuracy of the inner set of magnetic sensors might be compromised by various radiation effects and high heat loads, the complementary ex-vessel set is essential to provide backup information. This paper is an overview of the ex-vessel magnetic diagnostic which consists mainly of pick-up coils, steady state sensors, Rogowski coils in the toroidal field coil casing and fibre optic current sensors. The work presented aims at designing these sensors to meet the performance requirements in spite of the constraints due to the tokamak environment. The manufacturing constraints and the positioning requirements for all the ex-vessel magnetic sensors are described. The use and expected accuracy of the entire ex-vessel magnetic diagnostic is assessed in terms of magnetic equilibrium reconstruction and plasma current measurement precision.     

Design and analysis of multicavity prestressed concrete reactor vessels

A critical state-of-the-art review of multicavity prestressed concrete reactor vessel (PCRV) design and analysis practice is presented. Included are discussions of basic design concepts, the behavior of liners and penetrations, and the various tests required and/or employed to demonstrate acceptance of new vessel geometries and innovations. Brief reviews are given of the influences of design codes such as ACI/ASME Section III, Division 2, and BS4975; analysis methods including elastic, inelastic, and time-dependent techniques; the constituve equations that are essential to the satisfactory use of these techniques; and semi-empirical methods for calculating ultimate strengths of multicavity vessels.Tests conducted on liner plates, liner anchorage systems, and cooling tubes are reviewed together with the methods of analysis used in the design of anchorage systems. The adequacy and economy of present liner systems are considered and possible modifications in design are suggested.Design code requirements and methods of analysis for penetrations are discussed. The various types of closure designs that have been proposed and in some cases employed are evaluated on the basis of overall PCRV design philosophy. Several methods of prestressing PCRVs are considered with respect to relative advantages and disadvantages; existing overall vessel in-service inspection requirements are evaluated.     

Inelastic structural analysis: Design implications and experience

The current status of inelastic structural analysis is reviewed relative to the needs and requirements for design of elevated temperature nuclear reactor components. The classes of inelastic problems that can be solved in a practical way are identified and the overall scope of ongoing validation programs is reviewed. Three classes of problems are identified where the further development of simplified analysis techniques is necessary to make them tractable for design application.     

Design and analysis of a uranium hexafluoride actinide transmutation reactor

The long-term (> 1000 years) hazards of high-level wastes (HLW) can be reduced substantially by practising waste-actinide partitioning-transmutation (P-T). This paper investigates the waste-actinide transmutation performance of a uranium hexafluoride actinide transmutation reactor (UHATR). Using mostly present-day and near-term technology, a preliminary UHATR design is established. Because of the gaseous nature of the fuel, very high neutron fluxes are obtained. Compared with an LWR, the average blanket thermal flux of this UHATR is about 10–30 times higher, leading to a 15-fold improvement in the percentage of actinides fissioned per year of irradiation.     

Design and analysis of the W7-X divertor scraper element

Thehigh heat-flux divertor of the Wendelstein 7-X large stellarator experiment consists of 10 divertor units which are designed to carry a steady-state heat flux of 10 MW/m². However, the edge elements of this divertor are limited to only 5 MW/m², and may be overloaded in certain plasma scenarios. It is proposed to reduce this heat by placing an additional “scraper element” in each of the ten divertor locations. It will be constructed using carbon fiber composite (CFC) monoblock technology. The design of the monoblocks and the path of the cooling tubes must be optimized in order to survive the significant steady-state heat loads, provide adequate coverage for the existing divertor, be located within sub-millimeter accuracy, and take into account the boundaries to other in vessel components, all at a minimum cost. Computational fluid dynamics modeling has been performed to examine the thermal transfer through the monoblock swirl tube channels for the design of the monoblock orientation. An iterative physics modeling and computer aided design process is being performed to optimize the placement of the scraper element within the severe spatial restrictions.     

Design and analysis of direct action solenoid valve based on computational intelligence

Control Rod Hydraulic Drive Mechanism (CRHDM) is a newly invented patent of the Institute of Nuclear and New Energy Technology Tsinghua University which owns CRHDM's independent intellectual property rights while the integrated valve made up of three direct action solenoid valves is the key part of this mechanism. Therefore, the performance of the solenoid valve affects the integrated valve and the CRHDM directly. In this paper, we present a method to design the parameters of the direct action solenoid valve based on orthogonal experiment design, back propagation (BP) neural network and particle swarm optimization (PSO). The result proves that the method is feasible and accurate to design the parameters in order to obtain the biggest electromagnetic force. Besides, the result also shows that it is the current which influences the electromagnetic force of the direct action solenoid valve most.     

EAST装置通用数据分析软件的设计与实现

介绍了EAST 装置聚变实验数据分析与可视化集成软件EastScope的设计与实现.该软件采用ActiveX技术实现了多种格式数据访问方式的统一,调用了MATLAB引擎进行信号表达式运算,使用OpenGL技术完成了阵列信号的三维显示等.只需操作鼠标就可以完成信号的处理、分析、对比以及二维、三维显示,并能用交互式方式实现二维图形的任意轴缩放,三维图形的旋转、平移与缩放.     

Design of a magnetic spectrograph for surface, interface and thin-layer analysis

The design of a magnetic spectrograph of the QDQ type, to be coupled to a 6 MV Van de Graaff accelerator for the analysis with Rutherford backscattering and recoil spectroscopy of surfaces and thin layers, will be shown. Due to the high energy resolution of. the instrument (ΔE/E ≈ 2 × 10⁻⁴) the depth resolution is determined mainly by energy straggling and the combined effect of angular and lateral straggling, and is close to l monolayer near the surface. The mass discrimination of the spectrograph in combination with the energy dispersion of the detector in the focal plane makes possible the background-free detection of light (¹H−¹⁹F) atoms recoiling from a heavier substrate. The design of the total setup is such that channeling and channeling plus blocking experiments can be carried out, so that the position of light (or heavy) atoms on or in monocrystalline samples can be measured. The detector in the focal plane is position-sensitive in two dimensions so that with one setting of the spectrograph an energy range of 2% and an angular distribution of 5° (with a resolution of 0.1°) can be measured simultaneously. The opening angle in the energy-dispersive direction is 0.1°. The whole spectrograph, including the scattering chamber, can be rotated over 120° with respect to the beam line. For this purpose a bellows construction is made between the beam line and the scattering chamber, permitting the rotation while maintaining a vacuum of ≈ 10⁻¹⁰ Torr in the chamber.     

Design analysis of core assemblies for supercritical pressure conditions

The increase of steam parameters to supercritical conditions could reduce the power generating costs of light water reactors significantly [Proceedings of SCR-2000 (2000) 1]. Core assemblies, however, will differ from current BWR or PWR design. In this context, this paper summarizes the main results related to a thermal-hydraulic design analysis of applicable fuel assemblies. Starting from a thorough literature survey on heat transfer of supercritical fluids, the current status indicates a large deficiency in the prediction of the heat transfer coefficient under reactor prototypical conditions. For the thermal-hydraulic design of such fuel assemblies the sub-channel analysis code Sub-channel Thermal-hydraulic Analysis in Fuel Assemblies under Supercritical conditions (STAFAS) has been developed, which will have a higher numerical efficiency compared to the conventional sub-channel analysis codes. The effect of several design parameters on the thermal-hydraulic behaviour in sub-channels has been investigated. Based on the results achieved so far, two fuel assembly configurations are recommended for further design analysis, i.e. a tight square lattice and a semi-tight hexagonal lattice.     

Design failure mode and effect analysis for Korean fusion DEMO plant

The construction of Korean Fusion DEMO Plant (KFDP) for a demonstration of the plasma analysis and engineering feasibilities is planned in 2030s based on the Korean fusion technology roadmap. The radiation safety should be assured for nuclear facilities, so that, the KFDP is required to research for the regulatory requirements and industrial codes and standards. The final design guidance of the engineered safety features should be served in future. As the first step for this research, the failure modes and effects analysis in a design stage was performed. This leads to find the list of potential hazard elements and to obtain the list of initiating events for the future probabilistic risk assessment. The hazard elements expected to seriously threaten the integrity of the KFDP were investigated and determined to quantify the effect of the initiating events in the effect analysis: (1) a total loss of active cooling water to occur during the burn with decay heat calculation and (2) coolant ingress from cooling circuits into the vacuum vessel, cryostat and containment building. For those initiating events, the quantitative simulations using transient mass and energy calculation and computational fluid dynamics were performed.     

Design and analysis of concrete reactor vessels: New developments, problems and trends

This lecture reviews new developments in analysis and design of prestressed concrete reactor vessels (PCRV). After a brief assessment of the current status and experience, the advantages, disadvantages, and especially the safety features of PCRV, are discussed. Attention is then focused on the design of penetrations and openings, and on the design for high-temperature resistance — areas in which further developments are needed. Various possible designs for high-temperature exposure of concrete in a hypothetical accident are analyzed. Considered are not only PCRVs for gas-cooled reactors (GCR), but also guard vessels for liquid metal fast breeder reactors (LMFBR), for which designs mitigating the adverse effects of molten sodium, molten steel, and core melt are surveyed. Realistic analysis of these problems requires further development in the knowledge of material behavior and its mathematical modeling. Recent advances in the modeling of high-temperature response of concrete, including pore water transfer, pore pressure, creep and shrinkage are outlined. This is followed by a discussion of new developments in the analysis of cracking of concrete, where the need of switching from stress criteria to energy criteria for fracture is emphasized. The lecture concludes with a brief discussion of long-time behavior, the effect of aging, and probabilistic analysis of creep.     

Design and analysis of a thermal core for a high performance light water reactor

The High Performance Light Water Reactor is a Generation IV light water reactor concept, operated at a supercritical pressure of 25 MPa with a core outlet temperature of 500 °C. A thermal core design for this reactor has been worked out by a consortium of Euratom member states within the 6th European Framework Program. Aiming at peak cladding temperatures of less than 630 °C, including uncertainties and allowances for operation, the coolant is heated up in three steps with intermediate coolant mixing to eliminate hot streaks. Different from conventional reactors, the radial power profile is intended to be non-uniform, with the highest power in the first heat-up step in the core center and the lowest power in the second superheater step to result in the same peak cladding temperatures in each region. The concept has been studied with neutronic, thermal-hydraulic and structural analyses to assess its feasibility. Coupled neutronic/thermal-hydraulic analyses are defining the initial distribution of enrichment, control rod positions and the use of burnable poisons. Sub-channel analyses predict the coolant mixing inside assemblies, and a porous media approach simulates the flow of moderator water between assembly boxes. Finally, structural analyses of the assembly boxes are needed to minimize deformations during operation. Even though the core design cannot yet considered to be final, this state of the art review shall summarize the progress achieved so far and outline the remaining challenges.     

固态燃料熔盐堆自适应功率控制器设计及分析

钍基熔盐堆核能系统(Thorium-based Molten Salt Reactor,TMSR)是中国科学院首批启动实施的战略性先导科技专项,旨在研发第四代反应堆核能系统。固态燃料钍基熔盐实验堆(The Solid Fuel Thorium-based Molten Salt Experimental Reactor,TMSR-SF1)是一个10 MW热功率的氟盐冷却球床堆,目前已经完成方案设计和初步工程设计。功率控制系统是反应堆一个关键控制系统,实现反应堆正常启动、功率运行和正常停堆功能,对保证反应堆安全和稳定运行起着极其重要的作用。根据TMSR-SF1运行控制要求,结合自适应控制理论,基于Lyapunov稳定性理论设计了一种TMSR-SF1模型参考自适应功率控制器。基于TMSR仿真平台,使用MATLAB/Simulink建立了自适应功率控制系统模型,并开展了控制器特性分析。结果表明,自适应功率控制器具备良好的负荷跟随能力,抗干扰能力强、稳定性好、可靠性高,能够满足TMSR-SF1功率控制的要求,确保堆芯的输出功率与功率设定值相匹配。     

Design analysis of the hinge support for the ITER vacuum vessel

In order to verify design feasibility and structural integrity of a hinge type support for the ITER VV support system, the design analysis has been performed in detail, which includes heat transfer, elastic stress and limit analyses. The structural analyses were performed to confirm the transfer of forces through the supporting structure and to determine the maximum allowable loads according to the RCC-MR. From the heat transfer analysis for VV baking stage, total heat flow into the support was obtained to confirm the thermal heat flux into the cryostat under baking condition. In addition, the design modification was also discussed to enhance the structural performance of the supporting system.     

无损检测用加速器束流稳定性分析系统的研制

介绍了单道脉冲幅度分析系统对电子直线加速器每个宏脉冲内产生的电子束流稳定性的实时测量方法和对打靶产生的X射线剂量率进行动态监测的分析处理系统的研制与开发。介绍了单道分析器的原理和硬件组成。后端软件使用面向对象的Visual C^ .net开发完成，软件采用图形化软面板，所有控制和结果显示在计算机屏幕上。系统具有可靠性好、精度高、测量直观、应用软件的界面友好等优点。