Distribution of temperature rise in biological shield due to thermal neutrons

The aim of this work is the evaluation and calculation of the heat generation and temperature rise in local ordinary concrete due to capture of thermal neutrons. The total thermal and reactor thermal neutron fluxes were measured, and the results were used to evaluate the heat generation and temperature rise in ordinary concrete. In addition, the computer code ANISN (VAX version) and neutron multigroup cross-section library EURLiB-4 were used to calculate the total thermal neutron fluxes and the heat generation and hence the temperature rise. The results were displayed in curves to show the distribution of thermal neutron fluxes and the heat generation as well as the temperature rise in the shield thickness. The results showed a good agreement between the measured and calculated values. The results showed also that the heat generation and the temperature rise have their maximum values in the first layers of the shield thickness.     

Direct energy generation from γ-radiation

In this paper theoretical and experimental calculation of direct heat generation from the penetration of 6.12 MeV photons energy through single layer concrete, steel and lead shields is described. Also, the direct heat generation for various thickness for single layer concrete shield with ducts of different diameters is calculated.     

Direct heat generation and corresponding energy saving in nuclear reactors due to optimum synthesis of shielding materials

In this paper the calculation of direct heat generation and energy savings due to the penetration of 1.37 and 2.75 MeV energy photons, emitted from a Na-24 radiation facility, through double layer shielding slabs of aluminium, steel and lead is described. A comparison is being made among six different shielding material combinations in order to assess the optimum shield related to the maximum energy captured due to γ-rays penetration through the combined shielding materials.     

Calculation of γ radiation converted to thermal energy in nuclear reactor shielding facilities

The heat generation within a shield can be a primary consideration in its design, especially when the shield is for high power reactor in which rather large temperature increases can be expected. The heating up of the shield itself can be utilized via a suitable flow of water on the outside of the said shield or, in the case of thermal shields, using the nuclear reactor coolant leading to energy saving in nuclear power plants. The thermal energy generation rate inside the shielding materials is calculated using the build up factor method for multi-energy photons emitted from disc geometry radiation source and a comparison has been made in order to estimate the percentage of energy captured, and therefore saved, inside the shielding layers. In this paper the use of build up factor is suggested for thermal energy estimation in shielding materials such as Al, Fe and Pb and comparison with the results produced from the application of MCNP-4A computer code. This is a continuation of the work done by Bakos [Bakos, G.C., 1995. Benchmark data for g-rays emitted by an Na-24 source penetrating stratified shielding slabs. IEEE Trans. Nucl. Sci. 42 (2), 61–65].     

Evidence of vapor shielding effect on heat flux loaded on flowing liquid lithium limiter in EAST

A lithium (Li) vapour layer was formed around a flowing liquid Li limiter to shield against the plasma incident power and reduce limiter heat flux in the EAST tokamak. The results revealed that after a plasma operation of a few seconds, the layer became clear, which indicated a strong Li emission with a decrease in the limiter surface temperature. This emission resulted in a dense vapour around the limiter, and Li ions moved along the magnetic field to form a green shielding layer on the limiter. The plasma heat flux loaded on the limiter, measured by the probe installed on the limiter, was approximately 52% lower than that detected by a fast-reciprocating probe at the same radial position without the limiter in EAST. Additionally, approximately 42% of the parallel heat flux was dissipated directly with the enhanced Li radiation in the discharge with the liquid metal infused trenches (LIMIT) limiter. This observation revealed that the Li vapour layer exhibited an excellent shielding effect to liquid Li on plasma heat flux, which is a possible benefit of liquid-plasma-facing components in future fusion devices.     

Vapor Shield Models for Fusion Reactors Plasma-Facing Components

Radiant heat flux is a dominant mechanism by which energy transfers from the high-temperature core plasma to the interior critical components of the fusion reactor, which result in surface ablation and sever damage to the components. A vapor layer develops at the surface and provides a self-shielding mechanism at the plasma-material interface. Two models for the energy transmission factor through the boundary layer were developed and incorporated in the electrothermal plasma capillary code to predict the effectiveness of these models in surface self-protection. The electrothermal plasma capillary discharge simulates the typical conditions of fusion reactors disruption and quench phase and has been shown to be an adequate technique to evaluate the erosion of plasma-facing component. First model treats the radiant heat transport as it is affected by the variation of the plasma opacity, in which the vapor shield efficiency depends on the plasma optical thickness and the mean plasma opacity. The second model defines the vapor shield by the ratio of the energy reaching the surface to the total radiant energy emitted by the plasma with the inclusion of the plasma kinetic energy. The code can predict the axial and temporal variation of the transmission factor at each time step and mesh point, and predicts the plasma parameters with the effectiveness of the vapor shield at the boundary layer. The code prediction with implementation of both models has been used to compare the results with earlier ones and with some experimental data. Code results are in good correlation with experimentally measured ablation data.     

Fusion reactor first-wall cooling for very high energy fluxes

A study has been made of the feasibility of a protective first-wall shield between the plasma and the containment vessel for early experimental controlled thermonuclear fusion machines. The proposed first-wall shield is a water-cooled array of thin-walled tubes designed to take very high local energy fluxes originating from the neutral beam injectors. Detailed computer calculations reveal that heat flux capabilities of 3300 W/cm² are possible with first-wall shield sections made up of tubes 1 m long of Ta-10W alloy (with tubes of 10 mm i.d. and tube wall thickness of 0.5 mm) with a structural safety factor of about four. Required pumping powers on the order of 1 MW/m² of first-wall area exposed to these high energy fluxes are predicted for flow in the non-boiling regime. If operation in the subcooled nucleate boiling regime can be achieved without oscillations or instabilities, the required pumping power is shown to decrease by about an order of magnitude.     

辐射屏数及密度对超导腔垂直测试杜瓦漏热的影响

超导腔垂直测试杜瓦为广口液氦容器,对绝热性能有较高要求,其内筒体采用高真空多层绝热,顶部盖板侧为多屏绝热。为减少杜瓦顶部盖板侧漏热,提出了变密度辐射屏方法,以盖板、辐射屏、液面和氦气为对象建立了导热、对流与辐射耦合换热的传热模型,并通过实验进行了验证。通过数值计算得到了辐射屏数对漏热的影响规律及最优辐射屏密度。结果表明：1） 实验测得的辐射屏温度与传热模型计算结果较为一致,平均相对偏差为8.37%,认为传热模型是合理的;2） 靠近液面处第1层辐射屏（屏1）与液面间氦气的格拉晓夫数Gr随屏1温度T₁的升高存在极大值（T₁=9.14 K,Gr=1.12×10¹⁴）,T₁超过35 K后Gr基本保持不变;3） 等间距分布时,辐射屏数大于11层后总漏热变化不明显,一定辐射屏数下,相邻两屏之间气体导热占主导地位,低温区域（靠近液面侧）至高温区域（靠近盖板侧）导热热流比重减小,辐射热流比重升高;4） 11层辐射屏数下,高温区域布置7层、低温区域布置4层的变密度辐射屏方式漏热最小,与等间距分布相比漏热可减少4%。     

TOPAZ-Ⅱ空间核反应堆电源辐射屏蔽优化措施影响分析

空间核反应堆辐射屏蔽可减弱中子与γ光子通量对仪器设备的辐照,温度是影响辐射屏蔽性能的重要因素。利用Fluent软件对TOPAZ-Ⅱ空间核反应堆电源辐射屏蔽在真空环境下的换热行为进行了数值模拟,提出了优化措施,并揭示了优化措施对其温度场的影响和其温度分布特性。研究结果表明：冷却剂管道及其管槽表面喷涂低发射率的涂层具有均匀辐射屏蔽温度场和降低其温度峰值的效果,最优化参数为管道表面涂层的发射率为0.1～0.3、管槽表面涂层的发射率为0.1,且管道表面喷涂低发射率涂层效果要优于管槽;在冷却剂管道、管槽间添加单层表面发射率为0.04～0.07的真空遮热板也可使辐射屏蔽温度场与温度峰值处于最优状态。     

C-ADS主加速器系统CM4低温恒温器的热负荷分析

为了优化CM4低温恒温器的结构型式,本文对低温恒温器热负荷来源进行了研究,包括低温绝热支撑(POST)、电流引线、高功率耦合器的热传导,80K冷屏的热辐射,超导腔、高功率耦合器的动态热负荷等。研究结果表明,CM4低温恒温器满足低热负荷的要求,为下一步低温恒温器的优化及运行调试提供了理论基础。     

The use of an electrothermal plasma gun to simulate the extremely high heat flux conditions of a tokamak disruption

Disruption damage conditions for future large tokamaks like ITER are nearly impossible to simulate on current tokamaks. The electrothermal plasma source SIRENS has been designed, constructed, and operated to produce high density (> 10²⁵/m³), low temperature (1–3 eV) plasma formed by the ablation of the insulator with currents of up to 100 kA (100 s pulse length) and energies up to 15 kJ. The source heat fluence (variable from 0.2 to 7 MJ/m²) is adequate for simulation of the thermal quench phase of plasma disruption in future fusion tokamaks. Different materials have been exposed to the high heat flux in SIRENS, where comparative erosion behavior was obtained. Vapor shield phenomena has been characterized for different materials, and the energy transmission factor through the shielding layer is obtained. The device is also equipped with a magnet capable of producing a parallel magnetic field (up to 16 T) over a 8 msec pulse length. The magnetic field is produced to decrease the turbulent energy transport through the vapor shield, which provides further reduction of surface erosion (magnetic vapor shield effect).     

EAST-NBI主低温冷凝泵辐射传热优化分析

中性束注入器（NBI）主低温冷凝泵为J字型内置式，其防辐射挡板结构设计需综合考虑结构参数对辐射传热量及抽速的影响。本文选用蒙特卡罗方法追踪每份热辐射，以求解辐射传热量，并利用控制变量法得出辐射传热量与防辐射挡板结构参数之间的关系。模拟结果表明，壁面发射率、防辐射挡板结构单元的高度对低温冷凝板的辐射传热量影响较大，而结构单元的边界形状对辐射传热量影响较小。研究结果将为EAST-NBI主低温冷凝泵的优化提供依据并可为类似J字型内置式低温冷凝泵的设计提供参考。     

Heat transfer study for ITER blanket shield block cooling design

Two simplified models were developed for the cooling design of ITER shield block. Moreover, a new model, circular cylinder centered in a square solid, was also adopted to estimate the temperature, where the effects of heat transfer coefficient and volumetric heat rate were separated and studied individually. After that, the impact of dimension on the heat transfer in the new model was studied by a series of numerical analyses. At the last part, a numerical steady-state thermal analysis of a typical full shield block (SB) was performed to verify these models. Comparisons of the results from numerical analysis with these models show that the difference is acceptable in the practical application. The methods can be used not only for the cooling design, but also to know about the heat transfer in the SB.     

EASTICRF天线法拉第屏蔽的热-结构分析

在EASTICRF天线中,法拉第屏蔽是ICRF天线中的一个非常重要的部件。实验时,它位于真空室内直接面对等离子体,将承受着很大的热负荷。基于EASTICRF天线法拉第屏蔽结构的安全性,本文利用有限元的方法,首先对热负荷最大的法拉第屏蔽冷却管道在不同水流速下进行热分析,考察在不同水流速工况下法拉第屏蔽冷却管道上的温度分布情况,再通过热 结构耦合方法对法拉第屏蔽冷却管道进行结构分析,了解法拉第屏蔽在不同水流速下的应力大小和分布情况,分析结果为未来EASTICRF天线实验提供理论指导。另外,对法拉第屏蔽冷却管道结构进行了优化改进,并对优化改进后的法拉第屏蔽冷却管道在相同工况下进行了热和热 结构分析,分析结果确定了优化改进后的法拉第屏蔽冷却管道结构的优越性,分析数据为未来法拉第屏蔽冷却管道的优化改进提供理论指导,分析方法为其他同类装置提供有益的参考。     

包气带中不同示踪源层条件下核素迁移对比试验

为研究不同示踪源层材料对水份运移和核素迁移的影响，开展了以石英砂和黄土分别为示踪源层载体时Sr、Nd、Ce在包气带中迁移的对比试验。试验历时470d，在两种源层条件下各取土芯样4次。试验结果表明：在喷淋强度为5mm/h、3h/d条件下，Nd和Ce无论在石英砂示踪源层还是黄土示踪源层中浓度峰位均没有明显向下迁移；对于Sr，在黄土示踪源层中，470d后其峰位向下迁移约15cm(按质心计算为10cm左右)，而从石英砂示踪源层中向下迁移不明显（按质心计算，迁移约2.7cm），只是峰位有些展宽。上述结果表明，极薄（7mm）的细石英砂层也能对非饱和入渗水产生明显屏流作用，使得核素从石英砂示踪源层中向下迁移速度减慢。     

Radiation Heat Transfer and Vapor Shielding in a Two-Dimensional Model of an Electrothermal Plasma Source

Electrothermal (ET) plasma discharges are emerging as valuable mechanisms for pellet injection in magnetic confinement fusion reactors. They have been shown to be capable of achieving the required pellet velocities and pellet launch frequencies required for edge localized mode control. Another advantage of ET plasma discharges is their ability to simulate fusion disruption events by depositing large heat fluxes on exposed materials. A deeper understanding of the heat transfer processes occurring in ET plasma discharges will aid in this particular application. ET plasma discharges involve the passage of high currents (order of tens of kA) along the axis of a narrow, cylindrical channel. As the current passes through the channel, radiant heat is transferred from the plasma core to the capillary wall. Ablated particles eventually fill the plasma channel and the partially ionized plasma is ejected. It is well known that the ablated material separating the plasma core from the ablating surface can act as a vapor shield and limit the radiation heat flux reaching the ablating surface. In this work, the results from a two-dimensional simulation model for ET plasma discharges are presented. The simulation of the plasma in a two-dimensional domain combined with the diffusion approximation for radiation heat transfer is shown to successfully simulate the effects of the vapor shield layer that develops inside these devices.     

研究堆内材料释热率测量装置研制

为了探究材料释热率在研究堆孔道内的轴向分布规律,以高通量工程试验堆（HFETR）G7孔道为例,设计一种材料释热率测量装置。通过数值模拟方法得到释热率测量装置及试验段在载荷作用下的应变分布云图,采用物理计算得到量热计校对桥和测量桥的温度参数,并利用本装置在G7孔道开展释热率测量试验。结果表明,该装置整体结构满足强度要求,试验段量热计之间需加装保护管;计算得出样品、校对桥和测量桥的温度低于材料熔点,装置满足热工要求;试验测得的释热率值随堆功率变化规律性强,且不同材料在不同能量等级的γ射线环境下,对γ的吸收性是有区别的。因此,本装置可以作为HFETR释热率测量工具,为确定不同材料在堆内释热率分布情况提供保障。      

CEFR热传输及发电能力分析

针对中国实验快堆（CEFR）发电能力问题,对CEFR 3条回路的输热能力及热电转换能力进行建模并分析计算,通过40%额定功率发电工况下的试验数据验证了所建模型的正确性。依据CEFR的实际情况,提出优化运行工况的改进措施,其效果得到了汽轮发电机组厂家数据的验证,并结合输热能力验证模型模拟CEFR满功率下的热传输及发电能力。结果表明：基于优化后的运行工况,CEFR热传输系统和热电转换系统可达到设计要求。     

Transient analytical temperature distributions in cylindrical packed beds volumetrically heated by radiogenic decay energy

Analytical solutions are presented for the problem of the transient distribution of fluid and solid phase temperatures in a packed, porous, cylindrical particle bed with constant thermophysical properties. The packed particle bed is volumetrically heated by radiogenic decay energy from fission products. Flowing through the particle bed by forced convection is a single-phase fluid, either subcooled liquid or superheated vapor. The dynamic response of the packed bed is for low Reynolds numbers. In this case the transient will develop through the packed bed slowly enough for interphase heat transfer to keep the fluid and solid phase temperatures from having large differences. The two-dimensional, time-dependent Modified Dispersion-Concentric Model (D-C model) is used in the analysis of this problem. The D-C model energy equations are solved using Green's function. The mathematical solution characteristics for the transient fluid and solid phase temperature distributions are presented for three different volumetric heat generation terms: two-dimensional, time-dependent; simplified two-dimensional, time-dependent; and two-dimensional, time-independent. Using the two time-dependent volumetric heat generation terms, a comparison is presented for the transient fluid and solid phase temperatures and the radioactive decay heat power coming from the fission products in the particles.     

密度锁内分层传热特性的初步探讨

通过可视化观察方法,对3种不同实验管内的流体分层传热特性进行实验研究,同时,建立传热计算模型,对密度锁内的传热机理进行分析.结果表明:密度锁内的分层工质自上而下分为:混合层、界面层、导热层.混合层内的传热以对流为主,其余两部分的传热以导热为主.对不同管径的研究表明,密度锁内的蜂窝通道能有效地抑制扰动作用,减小混合层的厚度,降低通过密度锁的热量传递.