Numerical simulation of the transient temperature distribution within the cladding material during annealing in a high vacuum furnace

A high vacuum cylindrical furnace has a load of cylindrical tubes supported by a carriage. The temperature distribution is of particular interest for the industrial annealing of sheaths for nuclear reactor fuel pin cladding. A numerical model is used to calculate the temperatures in the load as a function of time, during both heating and cooling. Theory and experiment agree very well considering the sources of error. Theoretical results were also obtained for the heat transferred between furnace, carriage and tubes, as a function of time. The results allow the design of a complicated high vacuum industrial furnace, and the design and optimization of its loads, to proceed purely by computation.     

干凝胶法制备空心玻璃微球的炉内成球过程分析

基于干凝胶粒子炉内成球过程的分解实验结果及各阶段中间产物的分析测试结果,通过对干凝胶法制备空心玻璃微球工艺的传热、传质和运动的过程分析,将干凝胶法制备空心玻璃微球炉内成球过程合理简化为吸热、封装、气泡形成及聚并、精炼、冷却5个阶段。吸热阶段的升温速率以及发泡剂的分解温度和发泡效率、精炼阶段的精炼时间和温度、冷却阶段的冷却速率是影响干凝胶法制备空心玻璃微球工艺和空心玻璃微球最终质量的关键因素。     

Analytical and experimental investigation of buckling of fast reactor heat exchanger tubes

Intermediate Heat Exchanger (IHX) tubes in fast reactors are long and slender, and can be subjected to compressive loads during an emergency shutdown which could lead to buckling. A special experimental rig was constructed at Risley to carry out full-scale tests. The effects of temperature, gridplate misalignment, and variation in geometric and material Properties on the critical buckling load of the IHX tubes were examined. The results have shown that the critical buckling load is very sensitive to initial geometric imperfections present in the tube. The predicted results agree very well with those obtained experimentally.     

Thermal analysis and optimization of the EAST ICRH antenna

The ion cyclotron resonance of frequency heating(ICRH) plays an important role in plasma heating.Two ICRH antennas were designed and applied on the EAST tokamak.In order to meet the requirement imposed by high-power and long-pulse operation of EAST in the future,an active cooling system is mandatory to be designed to remove the heat load deposited on the components.Thermal analyses for high heat-load components have been carried out,which presented clear temperature distribution on each component and provided the reference data to do the optimization.Meanwhile,heat pipes were designed to satisfy the high requirement imposed by a Faraday shield and lateral limiter.     

Scaling of passive condenser system for separate effect testing

The passive containment cooling system (PCCS) of the simplified boiling water reactor (SBWR) is a passive condenser system designed to remove energy from the containment for long term cooling period after a postulated reactor accident. Depending on pressure condition and noncondensable (NC) gas fraction in drywell (DW) and suppression pool (SP), three different modes are possible in the PCCS operation namely the forced flow, cyclic venting and complete condensation modes. The prototype SBWR has total of six condenser units with each unit consisting of hundreds of condenser tubes. Simulation of such prototype system is very expensive and complex. Hence a scaling analysis is used in designing an experimental model for the prototype PCCS condenser system. The motive for scaling is to achieve a homologous relationship between an experiment and the prototype which it represents. A scaling method for separate effect test facility is first presented. The design of the scaled test facility for PCCS condenser is then given. Data from the test facility are presented and scaling approach to relate the scaled test facility data to prototype is discussed.     

Influence of multiple jet cooling on the heat transfer and thermal stresses in DEMO divertor cooling finger

The divertor concept for DEMO fusion reactor is based on modular design cooled by multiple impinging jets. Such divertor should be able to withstand a surface heat flux of at least 10 MW/m² at an acceptable pumping power. To reduce the thermal loads the plasma-facing side of the divertor is build up of numerous small cooling fingers. Each cooling finger is cooled by an array of jets blowing through the holes on the steel cartridge.The size, number and arrangement of jets on the cartridge influences the heat transfer and pressure drop characteristics of the divertor. Five different cartridge designs are analyzed in the paper. The most critical parameters, such as structure temperature, heat removal ability, pressure drop, cooling efficiency and thermal stress loadings in the cooling finger are predicted for each cartridge design. A combined computational fluid dynamics and structural model was used to perform the necessary numerical analyses. The results have shown that the cartridge design with the best heat transfer and pressure drop characteristics is not also the most favorable choice from the point of view of minimum stress peaks.     

2D thermal analysis for heat transfer from casing to winding pack in JT-60SA TF coils

In the framework of the Broader Approach Agreement, Europe is involved in the design activities for the Japanese Tokamak JT-60SA, investigating, among several issues, the operation of the superconducting TF magnets and their subsystems, aimed at the reactor conceptual design definition. In particular, one of the main critical aspects to study is the heating of the conductor due to both direct component of energy deposited by neutrons and by secondary gamma generated during plasma operation, and heat generated by the radiation on casing and transferred to the winding pack. Indeed, the operating temperature and the relevant temperature margin (i.e. the operating safety margin) of the magnet will depend strongly on the heat loads and on the capability of the coolant to remove it. Furthermore, the heat power to the conductor will depend on several aspects, namely the thickness of insulating material, the mass flow rate of helium flowing in the conductors and its thermodynamic properties at operating conditions, and the layout of the superconductors constituting the winding. Moreover, a crucial aspect in the final design will be the presence and position of the casing cooling channels. In this paper a 2D sensitivity analysis of heat transfer from casing to winding pack with respect to cooling channels number and position is presented, based on the reference layout of the magnet. As a result, we evaluated the optimum number and positioning of cooling channels needed, as a trade-off between magnet operating limits and available cryogenic power and if, at limit, they could be even neglected in normal operation, keeping dwell-time within reasonable values.     

The effect of certain cycle parameters on the efficiency of a nuclear gas turbine unit

The effect is analyzed of the increase of the effective pressure ratio, the regeneration factors, the initial gas pressure, temperature of the fuel element cladding, hydraulic resistance of the gas circuit on the internal efficiency of a nuclear gas turbine unit, taking into account the characteristics of the active zone of the reactor. The results are given of the effect on the efficiency of a nuclear gas turbine unit (NGTU) of the intermediate heating and cooling of the gas. A possible circuit for a NTGU is discussed, with one intermediate heating and three-stage cooling of the gas;Translated from Atomnaya Énergiya, Vol. 20, No. 5, pp. 412–415, May, 1966.     

底部支撑的快堆主容器挠曲分析

底部支撑的快堆主容器是由３个圆筒、２个圆锥壳、２个形状复杂的封头、１个大旋塞、１个堆芯支承结构和２个支承环组成的复杂结构。为了确定主容器及其支撑在地震载荷下挠曲特性的静态模型,将其简化为由４个圆筒、５个圆锥壳、２个椭圆壳和一个圆盘构成的复合结构,并将作为分析模型。然后,用有限元方法,计算出在垂直载荷、横向剪切载荷、剪切／垂直载荷多种变化速率下复合结构的挠曲载荷和挠曲模式。对高温引起的材料所氏模量降     

Cryogenic system design for HIAF iLinac

A helium cryogenic system is designed by the Institute of Modern Physics,Chinese Academy of Sciences,to supply different cooling powers to the cryomodules of ion-Linac (iLinac) accelerator,which serves as the injector of the High Intensity Heavy-Ion Accelerator Facility project.The iLinac is a superconducting heavy-ion accelerator approximately 100 m long and contains 13cryomodules cooled by superfluid helium.This article describes the cryogenic system design of the iLinac accelerator.The requirements of the cryogenic system,such as cooling mode,refrigeration temperature,operating pressure and pressure stability,are introduced and described in detail.In addition,heat loads from different sources are analyzed and calculated quantitatively.An equivalent cooling capacity of 10 kW at 4.5 K was determined for the cryogenic system according to the total heat load.Furthermore,a system process design was conducted and analyzed in detail.Further,the system layout and the main equipment are presented.     

Thermo-hydraulic design verification of the neutral beam liner for the ITER vacuum vessel

The ITER vacuum vessel has upper, equatorial and lower port structures used for equipment installation, utility feedthroughs, vacuum pumping and access inside the vessel for maintenance. A neutral beam (NB) port of equatorial ports provides a path of neutral beam for plasma heating and current drive. An internal duct liner is built in the NB ports, and copper alloy panels are placed in the top and bottom of the liner to protect duct surface from high-power heat loads. Global NB liner models for the upper panel and the lower panel have been developed, and flow field and conjugate heat transfer analyses have been performed to find out pressure drop and heat transfer characteristics of the liner. Heat loads such as NB power, volumetric heating and surface heat flux are applied in the analyses for beam steering and misalignment conditions. For the upper panel, it is found that unbalanced flow distribution occurs in each flow path, and this causes poor heat transfer performance as well. In order to improve flow distribution and to reduce pressure losses, hydraulic analyses for modified cooling path schemes have been carried out, and design recommendations are made based on the analysis results. For the lower panel, local flow distributions and pressure drop values at each header and branch of the tube are obtained by applying design coolant flow rate. Together with the coolant flow field, temperature and heat transfer coefficient distributions are also acquired from the analyses. Based on the analysis results, it is concluded that the lower panel for the NB liner is relatively well designed even though the given heat loads are very severe.     

Prototyping of the Blanket Shield Module for the ITER EC H&CD Upper launcher

The design of the ITER Electron Cyclotron Heating and Current Drive (ECH&CD) Upper launcher is recently in the first of two final design phases. The first phase deals with the finalization of all FCS (First Confinement System) components as well as with specific design progress for the remaining In-vessel components.The most outstanding structural In-vessel component of an ECH&CD Upper launcher is the Blanket Shield Module (BSM) with the First Wall Panel (FWP). Both of them form the plasma facing part of the launcher, which has to meet strong demands on dissipation of nuclear heat loads and mechanical rigidity. Nuclear heat loads from 3 MW/m³ at the First Wall Panel’ surface, decaying down to a tenth in a distance of 0.5 m behind of it will affect the BSM and the FWP. Additional heating of maximum 0.5 MW/m² due to plasma radiation must be dissipated from the FWP.To guarantee save and homogenous removal of such extensive heat loads, the BSM is designed as a welded steel-case with specific cooling channels inside its wall structure. Attached to its face side is the FWP with a high-power cooling structure.Based on computational analysis the optimum cooling channel geometry has been investigated. Specific pre-prototype tests have been made and associated assembly parameters have been determined in order to identify optimum manufacturing processes and joining techniques, which guarantee a robust design with maximum geometrical accuracy.This paper describes the design, manufacturing and testing of a full-size mock-up of the BSM. The study was carried out in an industrial cooperation with MAN Diesel and Turbo SE.     

低温永磁波荡器过冷液氮冷却系统的设计

低温永磁波荡器(Cryogenic Permanent Magnet Undulator,CPMU)是目前插入件技术发展的主要方向之一,其利用一些永磁材料,如钕铁硼(Nd Fe B)或镨铁棚(Pr Fe B)的磁场性能在低温下明显高于室温的特性来提高波荡器性能和光源束流品质,工作温区为50-150 K,需要冷却系统的冷却。CPMU冷却系统主要包括过冷液氮冷却系统和磁体阵列冷却回路。本文介绍了上海光源(Shanghai Synchrotron Radiation Facility,SSRF)CPMU过冷液氮冷却系统的设计方案和设计参数,进行了系统主要热负载的分析;对冷却系统中关键设备之一的过冷换热器进行了设计,并计算分析了过冷氮流经CPMU冷却系统的全程阻力损失,为系统另一关键设备液氮泵的选型提供依据。对CPMU过冷液氮冷却系统进行的在线测试表明,该设计满足CPMU样机的冷却需求。     

Advanced thermal-hydraulic analysis of the ITER TF magnets performed with VENECIA code for 15 MA reference scenario

Normal operation of the ITER TF coils at 15 MA reference scenario is simulated with the use of the VENECIA code. The developed numerical model adopts a full scale quasi 3D approach for thermal hydraulic and thermal diffusion analysis of TF coils at the reference scenario with greatly variable heat loads from nuclear heating and Eddy/AC losses. The model implements latest heat load specifications and corrective changes in design of TFWP, TF case and their cryogenic circuits. For the first time the primary auxiliary cryogenic boxes (ACBs) are included in a common model to provide for the forced-flow cooling of the TF winding, TF case together with CS/OIS structures and PF supports.     

Thermal hydraulic investigations of intermediate heat exchanger in a pool-type fast breeder reactor

Intermediate heat exchanger (IHX) in a pool-type liquid metal cooled fast breeder reactor is an important heat exchanging component as it forms an intermediate boundary between the radioactive primary sodium in the pool and the non-radioactive secondary sodium in the steam generator (SG). The thermal loads during steady state and transient conditions impose thermal stresses on the heat exchanger tubes and on the shells which hold the tube bundle. Estimation of these thermal loads and achieving uniform temperature distribution in the tubes and shells by having uniform flow distributions are the major tasks of thermal hydraulic investigations of IHX. Through multi-dimensional thermal hydraulic investigations performed using commercially available computer codes such as PHOENICS, the flow and temperature distributions in the tubes and shells and in its secondary sodium inlet and outlet headers are obtained with and with out provisions of flow distribution devices. The effectiveness of these devices in achieving acceptably uniform flow and temperature distributions has been assessed and thermal loads on the tubes and shells for thermo mechanical analysis of the IHX have been defined. The predictions of the computational studies have been validated against simulated experiments.     

Stress analysis of a nuclear reactor grid plenum assembly and its application to the EBR-II

A structural model is presented to predict the stress and deformation of a nuclear reactor grid plenum assembly. The model consists of two circular grid plates on top of one another which are interconnected by a large number of tubes, a perforated cylindrical shell and a cylinder at the outer edge of the plates. During normal reactor operation and shutdown the grid plenum assembly is subjected to both external and internal loads, in which the external load is due to the dead weight of the subassemblies resting on the upper grid plate and the coolant pressure, while the internal load is due to the temperature gradient and irradiation swelling in the grid plates accumulated through the life of the reactor.The structure is divided into many annular regions implying an axisymmetric treatment. The solution of the structural system is obtained by solving the equilibrium equations of the individual tubes, plates and shells, and satisfying the compatibility condition at the boundary of each region. The swelling strain in the grid plates is treated analogously to that of thermal strain in the elastic region. The numerical results of the problem are obtained by means of a computer code which uses a matrix-inversion technique to solve the simultaneous algebraic equations involved.     

Thermal-hydraulic design and transient analysis of passive cooling system for CPR1000 spent fuel storage pool

This paper proposes a design of passive cooling system for CPR1000 spent fuel pool(SFP). Our design can effectively manage the SFP temperature not to exceed80 C. Then the transient analysis of the CPR1000 SFP with designed passive cooling system is carried out in station blackout(SBO) accident by the best-estimate thermal-hydraulic system code RELAP5. The simulation results show that to maintain the temperature of CPR1000 SFP under 80 C, the numbers of the SFP and air cooling heat exchangers tubes are 6627 and 19 086, respectively.The height difference between the bottom of the air cooling heat exchanger and the top of the SFP heat exchanger is3.8 m. The number of SFP heat exchanger tubes decreases as the height difference increases, while the number of the air cooling heat exchanger tubes increases. The transient analysis results show that after the SBO accident, a stable natural cooling circulation is established. The surface temperature of CPR1000 SFP increases continually until 80 C, which indicates that the design of the passive air cooling system for CPR1000 SFP is capable of removing the decay heat to maintain the temperature of the SFP around 80 C after losing the heat sink.     

北京正负电子对撞机重大改造工程中超导四极磁体传热与流动计算

超导四极(SCQ)磁体是北京正负电子对撞机重大改造工程(BEPCⅡ)的关键设备之一。本文对SCQ磁体恒温器进行稳定运行状态下传热和流动计算。计算得到了磁体在低温下的热负荷以及磁体恒温器内各组成部分的温度分布，并在此基础上，提出减小SCQ磁体热负荷的方法。比较计算了SCQ磁体采用超临界和过冷液氦两种冷却方式对磁体稳定运行的影响。     

NSRL-U7C双晶单色器热载影响与计算分析

在测试合肥国家同步辐射实验室(NSRL)U7C X射线吸收精细结构(XAFS)光束线双晶单色器(DCM)热载影响基础上,进行了单色器在无水冷却、辐射散热条件下热平衡动态过程分析,并运用有限元方法对第一晶体作了热-结构耦合计算,得到晶体热平衡所需时间、温度梯度分布及由此产生的热应力形变与面形误差等;同时对拟增设的普通底面水冷方式作了比较分析.     

EAST装置高功率宽带射频发射机系统研制

离子回旋波加热系统是EAST装置最重要的辅助加热工具,作为系统最核心的分系统之一,高功率射频发射机为加热等离子体提供射频波能量,对提高等离子体运行参数起着极为重要的作用。基于电路分析、传输线和波导谐振腔等相关工程理论,本文系统地总结了射频发射机系统高功率放大器输入输出回路、放大器级间匹配、寄生振荡抑制、腔体冷却等部分的设计原理和实现方法。在假负载上进行了系统测试,在设计频段内获得了1.5 MW的射频输出功率,测试结果表明系统达到了设计的技术指标。通过两轮EAST射频加热实验验证,发射机系统运行稳定可靠,满足射频加热等相关物理实验要求。