Recombination Effect as Component of Residual Defect in Silicon Surface Barrier Detector

The thermal conductivity of graphite components used as in-core components in high-temperature gascooled reactors (HTGRs) is reduced by neutron irradiation during reactor operation. The reduction in thermal conductivity is expected to be reversed by thermal annealing when the irradiated graphite component is heated above its original irradiation temperature. In this study, to develop an evaluation model for the thermal annealing effect on the thermal conductivity of IG-110 graphite for the HTGRs, the thermal annealing effect evaluated quantitatively at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. Moreover, the thermal conductivity of IG-110 graphite was calculated by using a modified thermal resistance model considering the thermal annealing effect. The following results were obtained. (1) The thermal annealing effect on the thermal conductivity of IG-110 graphite could be evaluated quantitatively and a thermal annealing model was developed based on the experimental results at irradiation temperatures of up to 1,200°C and neutron fluences of up to 1.5 dpa. (2) The thermal conductivities of IG-110 graphite calculated by using the modified thermal resistance model considering the thermal annealing effect showed good agreement with experimental measurements. This study has shown that it is possible to evaluate the annealed thermal conductivity of IG-110 graphite by using the modified thermal resistance model at irradiation temperatures of 550–1150°C and irradiation fluences of up to 1.5 dpa.     

Modelling of the HFR-EU1BIS experiment and thermomechanical evaluation

Numerical modelling has been successfully applied during the design and evaluation of the HFREU1bis HTR fuel pebble irradiation experiment conducted by the Joint Research Centre Institute for Energy (JRC) in the HFR Petten, The Netherlands. HFR-EU1bis contains 5 HTR fuel pebbles that are placed in a graphite shroud. This assembly is placed in a containment tube, which is again placed in another containment, which is in contact with the HFR cooling water.The experimental requirement is to maintain a central temperature of 1250 °C for all pebbles throughout the irradiation time of 250 effective full power days (efpd). This was achieved by tailoring the gas gaps in the sample holder such that the axial heat generation profile is compensated. Determining the dimensions of the components of the experiment has been done using a thermomechanical finite element model. A combination of numerical and analytical techniques has been applied to minimise calculation times. The thermomechanical finite element model includes the thermal influence of changing gas gap dimensions due to expansion. Heat transfer by radiation through the gas gap has been modelled as well.Results of the model have been compared with measured temperatures. After some adjustments of the model parameters, good agreement has been found between calculated and measured temperatures. Additionally, the influence of irradiation on graphite thermal conductivity and thermal expansion has been included in the model. Due to irradiation damage the thermal conductivity of graphite decreases and thermal expansion generally increases, leading to increasing thermal stresses. Due to burn-up however, the heat flux decreases during irradiation, which leads to a decrease in thermal stresses. To evaluate the influence of these competing mechanisms, the thermal stress evolution during irradiation has been calculated and will be discussed.     

Dimensional changes and thermal conductivity by annealing and its relation to the defect concentration and stored energy release of neutron-irradiated graphite

The dimensional changes and thermal conductivity with the annealing of fine-grained isotropic graphite IG-110U and ETP-10 irradiated to 0.02 and 0.25 dpa (1.38 x 10²³ and 1.92 x 10²⁴ n/m², E > 1MeV) at a design temperature of <200°C were studied. The irradiated graphite exhibited a small volume expansion and large degradation in thermal conductivity. Post-irradiation annealing experiments were carried out on dimensional changes and thermal conductivity up to 1700°C, and the results were analyzed in terms of changes in the defect concentration of graphite crystals. The rapid recovery of thermal conductivity observed below 200°C in the graphite irradiated to 0.02 dpa was attributed to the annihilation of Frenkel defects, whereas the recovery observed in both dimension and thermal conductivity above 200°C in the graphite irradiated to 0.02 dpa and 0.25 dpa was caused by the annihilation of small interstitial clusters of 4 ± 2 atoms. The role of large clusters of interstitials and vacancies in the changes to smaller dimension than pre-irradiation at high annealing temperatures are discussed. The temperature dependence of stored energy release was estimated from the changes in defect concentration calculated from the recovery of thermal conductivity.     

真空石墨加热器温度场数值模拟与分析

石墨加热器是测量球床堆芯等效导热系数实验的关键部件,加热器温度场对系统安全及数据准确性有重要影响。本文基于Fluent计算平台,分别采用DTRM模型、P1模型、ROSSELAND模型、DO模型对真空保护环境下的石墨加热器温度场进行数值模拟,确定适合真空保护石墨加热器温度场的计算方法并讨论石墨导热系数、表面发射率对温度场分布的影响。比较分析表明：DO模型计算得到的温度分布较为接近真实情况,导热系数小于35W/(m•K)时,最高温度对其敏感;导热系数大于35W/(m•K)时,其对加热体最高温度影响较小,最高温度较为稳定。     

Thermal conductivity of highly neutron-irradiated beryllium in nuclear fusion reactors

Beryllium will be used as a neutron multiplier in Helium Cooled Pebble Bed (HCPB) DEMO blankets. The beryllium thermal conductivity is determining the maximum pebble bed temperature and, therefore, is very important for blanket design. Different grades of beryllium discs were neutron-irradiated at temperatures between 343 and 673 K and at fluences up to 1.6 × 10²³ cm⁻². At lower irradiation temperatures a significant drop of the beryllium thermal conductivity occurs even after small neutron fluences. With increasing neutron fluence, further moderate decreases of the conductivity are observed. With increasing irradiation temperature, the thermal conductivity further decreases. If the thermal conductivity of the irradiated beryllium is known, the conductivity of irradiated beryllium pebble beds can be assessed using the model suggested in this study.     

新型多元掺杂面对等离子体碳基材料的性能研究

对方案和机理的研究 ,在国内相关单位采取优势互补的方法进行合作 ,基于理论和实验 ,得出了高强度、高热导、低溅射掺杂石墨和碳 碳 (C C)复合材料的最佳配方 ;合适的工艺路线和工艺参数 ;掺杂石墨热导率提高 4倍以上 ;化学溅射产额下降 1个数量级 ;抗辐照增强升华和热解吸均有明显改善 ;抗等离子体辐照综合性能亦有较大提高 ,经受 1 0 0多次的等离子体放电轰击无裂纹     

The effect of residual strain on the thermal conductivity of nuclear graphite

This report describes an investigation of the effect of residual strain caused by compressive pre-stressing on the thermal conductivity of nuclear graphite. The structural changes caused by the applied stress were investigated by measuring the alteration in bulk and nitrogen displacement densities. In the range of small residual strains, the thermal conductivity increases as do the bulk and nitrogen displacement densities. For larger values of the residual strain (after stressing up to ~90% of the fracture stress) the thermal conductivity decreases approximately to its original value, while the bulk and nitrogen displacement densities continue to increase. These observations are compatible with some degree of crack formation; however, other data do not entirely support this simple picture.     

Theoretical analysis of mass transfer and reaction in a porous medium applied to the gasification of graphite by water vapor

The theories of mass transfer and reaction in porous medium were applied to investigate the gasification of graphite by water vapor in reactor. Based on several logical assumptions, the gasification of graphite by water vapor was investigated. Different shapes of graphite were analysed: semi-infinite blocks, planks and cylinders. Analysis reveals that the reaction rate varies with the location, which can explain the intense gasification that occurs at upstream during experiments. The temperature dependence of the reaction rate expressed different ranges. At low temperatures, the reaction rates depend greatly on the reaction temperature, and the influence of shape is more pronounced. At high temperatures, the dependence weakens due to the limit of gas transfer, and the influence of shape disappears. The water vapor pressure at exterior surface decreases with temperature, but the pressure gradients (both inside and outside of porous media) increases with temperature. For the platelike and cylindrical porous media, the reaction rate decreases with the thickness or radius because of the increase in specific surface. When the penetrating depth of water vapor is larger than the half-thickness or radius of porous media, the water vapor would cumulate in porous media so that the pressure of reacting gas increases.     

Temperature and composition dependence of the high flux plasma sputtering yield of Cu-Li binary alloys

Because of its very high thermal conductivity, actively cooled copper is an attractive plasma-interactive material for long pulse fusion devices such as ETR and devices with very high wall power loadings, such as reversed-field pinched (RFPs) and the proposed compact ignition torus (CIT). Pure copper however, has an unacceptably low threshold energy for runaway self-sputtering. Low Z materials such as graphite and beryllium are not subject to runaway self-sputtering, but suffer from high light ion erosion rates and very nonuniform redeposition. It has been suggested that strongly segregating alloys such as Cu-Li might be used to provide a low-Z self-sustaining coating while maintaining the desirable redeposition, thermal and mechanical properties of the majority alloy component.High flux deuterium plasma sputtering and ion beam experiments have been performed on Cu-Li alloys to determine if the reduction in copper erosion previously predicted and observed in low flux ion beam experiments occurs at particle fluxes representative of an RFP first wall or tokamak limiter. Partial sputtering yields of the copper and lithium components have been measured as a function of alloy composition and sample temperature using optical plasma emission spectroscopy, weight loss and catcher foil techniques. It is found that the lithium sputtering yield increases with increasing sample temperature while the copper yield decreases by as much as two orders of magnitude. The temperature required to obtain the reduction in copper erosion is found to be a function of bulk lithium concentration. Consequences of these experimental results for anticipated erosion/redeposition properties are calculated, and the Cu-Li alloy is found to compare favorably with conventional low-Z materials.     

The preparation of fine-grain doped graphite and its properties

Fine-grain doped graphite was prepared by the ball-milling dispersion method for the first time. Such composite has not only high thermal conductivity and excellent bending strength (116 MPa), but also better oxidation resistance at elevated temperature and outgassing properties than those of composite doped with normal size carbides. Correlations between microstructure and properties of such composites are also discussed in detail.     

The thermal conductivity of hexagonal boron nitride parallel to the basal planes

Detailed measurements of the basal thermal conductivity and specific heat of boron nitride have recently been reported over the range 1.5–350 K. Boron nitride is a lattice conductor with a conductivity varying as T^2.4 below 10 K, while the lattice specific heat varies as T³ over the same range. This anomalous difference in the temperature dependence of lattice conductivity and specific heat is similar to that observed in graphite. The theory of the lattice conductivity of graphite parallel to the basal planes is applied to the explanation of this anomaly. It appears that the value of C₄₄ in this lattice must be smaller than that of graphite.     

The loosing mechanism of screw bolts on the first wall graphite tiles in EAST

The first wall in Experimental Advanced Superconducting Tokamak (EAST) has used many graphite tiles to face the plasma. All graphite ties have been mounted on heat sink using screw bolts which had been preloaded to produce clamp force. The clamp force is very important to keep the graphite tiles on the surface of the heat sink tightly because the heat flux will cross this contacting surface in a small thermal resistor. Without the clamp force the small gap between graphite tiles and heat sink will lead thermal power cannot be carried away by cooling water. The worse is some bolts turned out with the loss of clamp force. From the mathematical models the process of the loss of clamp force has been studied. Research results explain how the different thermal expansion of three members of the screw joint leads to the clamp force decrease to zero under temperature rise and external force and how the stiffness affect the relationship between the clamp force and temperature. The research also gets the critical temperature point that the clamp force can remain above zero. Research results indicated this screw joint is absolute to lose their clamp force during the EAST running so that the bolt joints should be redesign to improve its reliability.     

基于在线耦合的球形燃料元件温度场计算

由于三层各向同性(TRISO)颗粒弥散型燃料元件结构复杂且其材料性能随着辐照水平不断变化，不同燃耗下燃料元件的等效热导率不易确定。本研究基于COMSOL软件完成了TRISO颗粒性能分析程序开发，并与BISON程序预测值进行了对比分析。随后，基于COMSOL软件与MATLAB联合仿真建立了球形燃料元件等效热导率的计算方法，实现了球形燃料元件和TRISO颗粒模型间的在线耦合计算。在此基础上，获得了不同边界温度、燃耗条件下燃料元件径向等效热导率分布及温度场分布。计算结果表明，快中子注量达到3×10²⁵m^–2时，TRISO等效导热率下降约20%，燃料等效热导率下降约15 W/(m·K)。为了验证本研究方法的有效性，用微分-有效介质理论模型(D-EMT)计算燃料的等效导热率，得到的球形燃料中心温度预测值相比本研究方法的预测值低约25 K。本文研究方法更能真实反映球形燃料元件在反应堆内的温度场变化。     

Analytical Modelling of Overpack of Enriched Uranium Hexafluoride Package in a Fire Event

Abstract

A fissile package must be subjected to a fire test at 800°C for 30 min. For a package that contains enriched uranium hexafluoride (UF₆), the temperature of the contents has to be kept below 121°C in order to avoid rupture of the 30B cylinder. A previous study on heat transfer mechanisms in an enriched UF₆ package type ‘DOT-21PF-1’ found that decomposed gas with a temperature of around 100°C generated from heated penolic foam thermal insulator was a major heat source heating the cylinder in the initial period of the fire test, in addition to the heat conducted from the outer surface through the thermal insulator into the interior of the package. Mitsubishi Materiass Corporation has developed a new packaging for enriched UF₆ ‘MST-30’ that also uses phenolic foam as the thermal insulator, and has pelformdd a fire test at 800°C for 30 min, which also indicated the temperature history of the package interior affected by heating from the decomposed gas. In this study, a laboratory test was carried out to identify the kind of gas generated from the heated phenoiic foam. The test found that the decomposed gas was water vapour. Since water vapour has a large latent heat, it has a large heat transfer coefficient when it condenses on a cool surface. By taking the water vapour heating mechanssm into account, an analytical model was developed for MST-30 to simulate the measured cylinder temperature history. Analytical models for evaluaiing the cylinder temperature under the 800°C/30 min fire test considering conservativeness are also presented.     

Modeling of void, fast power and graphite temperature reactivity coefficients measurements for the validation of RELAP5-3D RBMK-1500 reactor model

This paper deals with the modeling of RBMK-1500 specific transients taking place at Ignalina NPP: measurements of void and fast power reactivity coefficients, as well as change of graphite cooling conditions transient. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and based on the obtained experimental results the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is unique and important from the point of view of model validation for the gap between fuel channel and the graphite bricks. The measurement results, obtained during this transient, enabled to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors.     

Tritium release behavior from the graphite tiles used at the dome unit of the W-shaped divertor region in JT-60U

Release behavior of tritium from the graphite tiles used at dome top and inner dome wing in JT-60U was investigated by the thermal desorption method in dry argon, argon with oxygen and water vapor, or argon with hydrogen. It was found that approximately 20-40% of total tritium is left in graphite even after heating to the high temperature above 1000 °C in dry argon. The residual tritium could be removed by exposing the graphite tile to oxygen with water vapor or hydrogen at the high temperature above 1000 °C. The tritium retention of the dome top tile was quantified as 84-30 kBq/cm². The inner dome wing tile had a steep tritium distribution from 8 to 0.1 kBq/cm². It is observed that a measurable amount of tritium existed in the deep site of the graphite tile.     

Transient gasification in an NBG-18 coolant channel of a VHTR prismatic fuel element

This paper investigates the transient gasification of NBG-18 nuclear graphite with atmospheric air ingress in a 0.8-m long coolant channel of a prismatic Very High Temperature Reactor fuel element. Analysis varied the initial graphite and air inlet temperature, T_o, from 800 to 1100 K at air inlet Reynolds number, Re_in = 5, 10 and 20. The analysis employs a Generic Interface that couples a multi-species diffusion and flow model to readout tables of the CO and CO₂ production fluxes. These fluxes are functions of the graphite local surface temperature, oxygen partial pressure and graphite weight loss fraction and calculated using a chemical-reactions kinetics model for the gasification of nuclear graphite. The analysis accounts for the heats of formation of CO and CO₂ gases, the heat conduction in the graphite sleeve, and the change in the oxygen partial pressure in the bulk gas flow mixture along the channel. Transient calculations performed up to a weight loss fraction of 0.10 at the entrance of the channel, t₁₀. They include the local graphite surface temperature and composition of bulk gas flow, the local and total graphite weight losses and the local and total production rates of CO and CO₂ gases. The heat released in the exothermic production reactions of these gases increases the local graphite surface temperature, accelerating its gasification. At the end of the calculated gasification transient, t = t₁₀, the graphite weight loss is highest at the channel entrance and decreases rapidly with axial distance into the channel, to its lowest value where oxygen in the bulk gas flow is depleted. Increasing T_o decreases t₁₀ and the total graphite loss, while increasing Re_in decreases t₁₀ but increases graphite loss.     

重水研究堆热柱石墨潜能释放分析研究

重水研究堆堆内石墨构件在长期中子辐照下将会累积潜能,为确保重水研究堆堆内石墨构件安全退役及处理处置,本文采用差示扫描量热仪对重水研究堆3个不同位置所取热柱石墨样品进行了潜能测量,扫描温度范围为10～550 ℃、升温速率为10 ℃/min。结果表明：3个位置的样品在80～500 ℃温度积分区间内潜能释放量分别为70.690、42.167、18.158 J/g;潜能释放率曲线峰值温度均大于300 ℃,未辐照石墨样品的比热容较热柱石墨样品释放率dS/dT（S为潜能释放量（J/g）,T为温度（℃））高,表明本实验所取石墨样品不会发生潜能释放导致石墨自身温度上升的情况;3个位置样品的快中子注量分别为6.75×10¹⁶、6.10×10¹⁴、1.89×10⁷ cm^-2;获得了潜能释放分数曲线与潜能释放速率曲线,1#和2#位置样品的潜能释放速率曲线具有至少2个释放峰,表明潜能释放过程中具有至少2个动力学过程。     

Immobilization of Radioactive Wastes in Hydrothermal Synthetic Rock, (III)

The high-level radioactive waste form (21.8w/o waste, 47.7w/o α-quartz, 20.5w/o amorphous aluminosilicate, 10w/o Al(OH)₃) was produced with the addition of 10-N NaOH solution under hydrothermal hot-pressing conditions at 350°C and 66 MPa for 6 h. In order to characterize the waste form, following tests and measurements were performed: Soxhlet flow leach tests (97°C), static leach tests in deionized water at low temperatures (40, 90°C) and under hydrothermal conditions (100–350°C), crystalline phase determination, microstructure observation, compressive strength measurement, weight loss measurement by heat treatment and thermal conductivity measurement.

The waste form was mainly composed of α-quartz and had porous structure. As for leach rates determined by static low temperature leach tests, the leach rates of the waste form were much lower than a concrete form but slightly higher than a glass form. The waste form was stable under hydrothermal conditions in comparison with a glass form. It had high compressive strength and thermal resistance. Its thermal conductivity was higher at room temperature than that of a glass form.     

The effective thermal conductivity of crud and heat transfer from crud-coated PWR fuel

Water-filled crud on the surface of PWR fuel could offer resistance to the flow of heat, which might be expected to cause higher clad temperatures, and probably more fuel failures, than are actually observed. However, there is some evidence from post-irradiation inspection that the crud is penetrated by pores large enough to permit vapour formation, and it is believed these provide a mechanism for ‘wick boiling’ to occur, which modifies, and indeed can under some circumstances actually improve, heat transfer. This phenomenon is investigated using a two-dimensional coupled multi-physics model, accounting for the flow of water, heat and dissolved species within the crud. The fuel thermal performance is characterized in terms of an effective crud thermal conductivity derived from the use of this model, and the non-linear dependence this effective thermal conductivity has on parameters such as crud thickness and pore density is determined.