On the analogy between thermally and irradiation induced creep

Employing an analogy between thermally induced and irradiation induced creep, physical arguments are used first to deduce a one-dimensional constitutive relation for metals under stress in a high temperature and high neutron flux field. This constitutive relation contains modified superposition integrals in which the temperature and flux dependence of the material parameters is included via the use of two reduced time scales; linear elastic, thermal expansion and swelling terms are also included. A systematic development based on thermodynamics, with the stress, temperature increment and defect density increment as independent variables in the Gibbs free energy, is then employed to obtain general three-dimensional memory integrals for strain; the entropy and coupled energy equation are also obtained. Modified superposition integrals similar to those previously obtained by physical argument are then obtained by substituting special functions into the results of the thermodynamic analysis, and the special case of an isotropic stress power law is examined in detail.     

Research reactor power controller design using an output feedback nonlinear receding horizon control method

A constrained, output feedback nonlinear receding horizon control (NRHC) method is applied to design a research reactor power controller. The method uses a nonlinear plant model subject to state, control and terminal set constraints; a nonlinear cost function; and a high gain observer. The controller regulates reactor power from 1% to 100% of full power; considers known disturbances, such as reactivity insertions and changes in core inlet flow and temperature; and includes upper limits constraints on neutron flux, neutron flux rate, core outlet temperature and core inlet–outlet temperature difference. Simulation results show an excellent performance for power regulation and known disturbances rejection: all process variables are kept within the admissible limits avoiding the actuation of the safety systems.     

Irradiation temperature, flux and spectrum effects

Although great progress has been made in understanding the irradiation behaviour of reactor pressure vessel (RPV) steels, many aspects are still not fully understood. A large amount of data has been generated for understanding the effects of different irradiation conditions on material properties. The data needed for the long term operation of RPVs is almost always created by accelerated irradiations in test reactors, and due to insufficient knowledge on the damage interaction between the material and the high energy neutrons the potential bias of the conclusions on material properties in non-accelerated irradiation conditions can not be excluded. Important parameters for the extrapolation of results from accelerated irradiations to typical power irradiation conditions are the irradiation temperature, the neutron flux and the neutron spectrum. In particular, the effect of neutron flux on embrittlement behaviour is considered a complex phenomenon, and it seems to be dependent on the alloy composition, the neutron fluence range and the irradiation temperature. This paper will present the current knowledge on temperature, flux and spectrum effects, based on a recent literature survey and other relevant publications on the subject. It will explore the implications these effects may have for the safety evaluation of aged RPVs, especially for those exposed to long irradiation periods.     

Slow Neutron Scattering and Classical Mean-Square Displacement Functions of Atoms

The resonance absorption integral and its temperature coefficient for finite ²³⁸UO₂ rods with a density of 10g/cm³, radii of 0.2, 0.3, 0.5, 1.0 and 2.0cm, and heights of 10, 20 and 50cm are calculated by using the RICM code and collision probabilities derived by improving Carlvik's equation.

It is found that assumption of infinite rod height underestimates the resonance integral and the coefficient by about 4.5 and 2.5%, respectively, in the case where the height is 10cm and the radius 2 cm, if the finite cylinder is substituted by an infinite one of the same radius. By adopting an infinite cylinder with the same mean chord length as the finite one, however, substitution of the finite by infinite cylinder is not unduly affect accuracy.

Next, the validity of the flat flux approximation in the axial direction of a finite cylinder is examined by using collision probabilities for a cylinder system divided along its axis into meshes of equal height. It is concluded that no correction is necessary for this approximation.     

Effects of Non-uniform Moderator Temperature on Core Power Distribution in Accelerator Driven Reactors

In a thermal reactor with moderators at different temperatures, a difference arises in the average speeds of thermalized neutrons between the high temperature part and the low temperature part of the moderator, and the non-uniformity of the average speed of thermalized neutrons may effect changes in the spatial dependence of the thermal neutron flux in a core. To investigate the thermal neutron flux in the case the average speed of thermalized neutrons is dependent on the position within a core, time-dependent two-group diffusion equations were applied. The influence of a nonuniform moderator temperature on the core power distribution was investigated about a graphite-moderated subcritical reactor driven by periodic injections of pulsed fast neutrons. The cylindrical reactor model by which a high temperature part of a core that has a spallation target at the center is enclosed by a low temperature part of a core was used. Changes in the core power distribution were calculated. It turned out that the momentary increases of a thermal power density caused by periodic injections of pulsed fast neutrons increase as the difference in the average speeds of thermalized neutrons in the high temperature part and the low temperature part of a core increases.     

Evaluation of The Thermal Performance of Multi—Element Doped Graphite under Steady—State High Heat Flux

Multi-element doped graphite,GBST1308 has been developed as a plasma facing material(PFM) for high heat flux components of the HT-7U device.The thermal performance of the material under steady-state(SS) high heat flux was evaluated under actively cooling conditions,the specimens were mechanically joined to copper heat sink with supercarbon sheet as a compliant layer between the interfaces.The experiments have been performed in a facility of ACT (actively cooling test stand) with a 100kW electron gun in order to test the suitability and the loading limit of such materials.The surface temperature and bulk temperature distribtuion of the specimens were investigated.The experimental results are very encouraging that when heat flux is not more than 6 MW/m^2,the surface temperature of GBST1308 is less than 1000℃,which is the lowest,compared with IG-430U and even with CX-2002U(CFC),The primary results indicate that the mechanically-joined material system by such a proper design as thin tile.Super compliant layer,GBST as PFM and copper-alloy heat sink,can be used as divertor plater for HT-7U in the first phase.     

Ultimate analysis of BWR Mark III reinforced concrete containment subjected to internal pressure

Numerical analyses are carried out by using the ABAQUS finite element program to predict the ultimate pressure capacity and the failure mode of the BWR Mark III reinforced concrete containment at Kuosheng Nuclear Power Plant, Taiwan, R.O.C. Material nonlinearity such as concrete cracking, tension stiffening, shear retention, concrete plasticity, yielding of reinforcing steel, yielding of liner plate and degradation of material properties as a result of high temperature effects are all simulated with proper constitutive models. Geometric nonlinearity as a result of finite deformation has also been considered. The results of the analysis show that when the reinforced concrete containment fails, extensive cracks take place at the apex of the dome, the intersection of the dome and the cylinder and the lower part of cylinder where there is a discontinuity in the thickness of the containment. In addition, the ultimate pressure capacity of the containment is 23.9 psi and is about 59% higher than the design pressure 15 psi.     

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.     

V型凹口对液态熔盐堆用核石墨构件应力分析的影响

本文利用了一个根据球床模块堆(Pebble Bed Modular Reactor,PBMR)用核石墨材料辐照性能数据编写的用户自定义材料模型(User defined Material model,UMAT),按照美国橡树岭国家实验室(Oak Ridge National Laboratory,ORNL)的液态燃料熔盐试验堆(Molten Salt Reactor Experiment,MSRE)用核石墨构件尺寸,为钍基熔盐堆(Thorium-based Molten Salt Reactor,TMSR)设计了一款方型核石墨构件。利用新编UMAT对该核石墨构件进行了初步的应力分析。分析结果表明,在没有预制裂纹的情况下辐照梯度越大核石墨构件中心区域最大主应力值越大,构件的断裂位置可能出现在构件中心位置处;对于有V型凹口预制裂纹的情况,应力集中部位均出现在预制裂纹尖端附近,这将可能导致裂纹尖端附近出现裂纹扩展,从而引起构件断裂失效。     

承压热冲击下DVI接管动态应力特性研究

压力容器直接注入（DVI）接管在热冲击下的动态应力特性对于反应堆压力容器（RPV）结构完整性评估具有重要意义。建立了含DVI接管的RPV压力壳热流固耦合数值计算模型,并进行了验证分析;然后研究了蓄压安注箱（ACC）和堆芯补水箱（CMT）安注时RPV筒体和DVI接管热工水力特性;最后分析了热冲击下RPV筒体和DVI接管连接高应力区的温度分布、等效应力和等效塑性应变分布特性。研究结果表明,ACC安注阶段RPV筒体和DVI接管连接区存在较大的温度梯度和等效应力,且发生了局部塑性变形。若发生承压热冲击事件,应控制好DVI接管连接区温差,确保反应堆压力容器的结构完整性。本文开发的热冲击下热流固耦合数值计算模型和计算方法可用于核岛内DVI接管与RPV筒体的安全性评价,也可用于类似承压结构在热冲击下的动态应力特性分析。      

针对球床的三维特征线方法几何建模研究

特征线方法通过在计算区域密置特征线来计算角通量,对于计算区域的材料分布和几何结构没有要求,因此特征线方法的几何处理能力受制于几何描述模块对于各种几何区域的描述能力。基于体素构造（CSG）方法,开发了三维特征线程序MOCP的几何描述模块。该几何描述模块可描述随机分布的球床。针对球形燃料的网格划分方式进行了研究,临界球的计算结果表明,当径向网格超过30层时,k_eff的相对误差小于0.1%。通过对几何描述方式的改进大幅提高了三维特征线追踪的效率,并且实现了在各种形状边界上的特征线布置。     

Material Properties of a Natural UF6 Transport Cylinder Vessel at High Temperature

Abstract

The IAEA Regulations for the Safe Transport of Radioactive Material are to be revised in 1996 and the fire test (800°C for 30 min) could become a requirement for the natural UF₆ transport cylinder. ASME SA 516 carbon steel is used as the structural material for this type of cylinder. It is very important to obtain high temperature data for SA 516 steel to be able to evaluate the integrity of the UF₆ transport cylinder vessel in the fire test. CRIEPI has therefore conducted material tests on SA 516 at high temperatures. The AC₁ and AC₃ transformation points of actual SA 516 steels have been measured. Tensile tests up to 900°C were conducted using USA, French and Japanese manufactured materials and the influence of phase transformation assessed. Preliminary creep tests show that assessment by creep strength can give a more conservative estimation than using the tensile strength. Creep deformation equations have been obtained using uniaxial creep tests and internal pressure creep tests. In addition, by the use of internal pressure creep rupture tests, the relation between the circumferential stress, the test temperature and the rupture time has been obtained.     

Safety analysis for an inadvertent control rod withdrawal event during a power operation of an advanced integral reactor

To identify a safety margin in the case of an inadvertent control rod withdrawal event of a 65-MW_t advanced integral reactor, safety analysis has been carried out by using the Transients And Setpoint Simulation/System integrated Modular Reactor (TASS/SMR) code. The diverse initial conditions, various reactivity insertion rates into a core, different combinations of a reactivity feedback and three different speed modes of a main coolant pump (MCP) have been considered to identify the effect of each parameter on a critical heat flux ratio (CHFR) and the initial condition resulting in the worst consequences from the viewpoint of the minimum critical heat flux ratio. The analysis results show that the worst consequences occur when a reactivity of 17.61 pcm/s is inserted into a core at an initial condition of a 45% initial core power, high coolant temperature at the core inlet position, low system pressure and a thermal design flow. It is also assumed that the least negative fuel and moderator temperature coefficients are applied. The safety parameters such as the minimum critical heat flux ratio and the system pressure are maintained within the safety limits and the reactor is safely transferred to a safe condition by a functioning of the safety systems of the advanced integral reactor.     

Transient heat conduction problems in inhomogeneous media discretized by means of boundary-volume element

An integral equation formulation is presented for the transient heat conduction problems in inhomogeneous media. The material constants are assumed to be prescribed as arbitrary, continuous and differentiable functions of position vector. The governing integral equations are derived from the weighted residual statement of the problems in which the fundamental solution to the corresponding heat conduction problems in homogeneous media is used as the weight function. The whole domain of interest is discretized into a series of boundary-volume-time elements, and then a set of linear simultaneous equations are obtained. Their solutions yield the temperature in the whole domain as well as the heat flux on the boundary.     

A structural analysis of the circular cylinder with multi holes under thermal loading

A structural analysis of the circular cylinder with multi holes is performed using the finite element analysis program . This structure is an analytical model of the capsule used for material irradiation tests. The temperature distributions of the cylinder due to gamma heating are obtained and various parameters, such as specimen size, quantity of specimens and gap sizes between the holder and the specimen are considered in the analysis to obtain the thermal and mechanical characteristics. To assess the structural integrity of the capsule, stress analysis under thermal loading is also performed. The analysis results show that, in all specimens, the peak temperature occurred, and is significantly dependent on gap sizes between the holder and the external tube or the specimen. The stress of the cylinder, under thermal loading, is lower than the allowable stress of the material used.     

Kinetics models for safety studies of accelerator driven systems

Safety analyses of Accelerator Driven Systems (ADSs) are mainly performed by codes developed in the past for critical reactors, a point-kinetics model for computing the transient power being often employed. It is shown that this model – that assumes time-independent neutron direct and adjoint flux shapes – may be inaccurate in the ADS case even if it is acceptable for a similar “critical” case. Although the material distribution remains almost unchanged, flux (power) shape variations may be significant in the first case due to external source related effects. An option for a more refined modelling of the neutron adjoint flux in ADS analyses is discussed. This option is shown to be rather complicated in the general case: it may give rise to involvement of several adjoint flux shapes and several sets of related point-kinetics parameters (reactivity, etc.). To improve the accuracy of the point-kinetics treatment, an extended point-kinetics model, which employs several flux (or power) shapes precomputed at steady-state conditions, is proposed in the paper. This approach may help to avoid the spatial kinetics treatment if no strong material movement occurs. The reactivity and other point-kinetics parameters are defined similarly to a critical reactor.     

Dissipative heating in a coated viscoelastic cylinder

The thermomechanical responses of a long coated hollow cylinder made of viscoelastic material subjected to cyclic presure fluctuations at high frequency are studied. The accumulation of the continuous energy dissipation induces a significant temperature increase in the material which affects the material behavior and hence the structural response. The paper adopts a numerical approach with the intention of studying the effect of coating. It is found that the coating has little effect on the dissipative heating due to internal pressure fluctuation. However, the heating effect can be greatly reduced when the cyclic pressure is applied on the outside surface.     

Irradiation Test of Stainless Steel by High Power CO2 Laser

Electron beam, plasma arc and ion beam are often employed to simulate the high heat flux applied to the first wall or the divertor plate in a fusion reactor. In this study, an irradiation test with high heat flux was carried out under atmospheric condition by using high power CO₂ laser. The test material is SUS316 and the temperature change and the melting amount were measured. A thermal analysis code to take melting and evaporation behavior into account was developed. The laser absorption coefficient can be raised up to 95% before melting if special paint is coated on specimen surface. After melting, this coefficient is estimated to be 60% by thermal analysis. However, it was revealed that a precise modification of this model was indispensable. Although the effect of irradiation environment or heat source on material damage was also examined, there is no significant difference one another. In conclusion, it is found that CO₂ laser is quite suitable for use as a heat source to simulate a high heat flux.     

新型耐高温中子屏蔽复合材料的设计制备及性能研究

采用表面改性处理技术,制备了由环氧树脂、B₄C（或BN）和聚丙烯酸铅组成的新型耐高温中子屏蔽复合材料,重点研究了材料制备工艺及主要性能指标,利用蒙特卡罗程序MCNP对材料中子屏蔽性能进行了模拟计算,并与文献报道的屏蔽材料铅硼聚乙烯进行了比较。结果显示,由环氧树脂、B₄C和聚丙烯酸铅组成的复合材料各项力学性能良好,具有良好的耐高温性能,210 ℃烘烤7 h外观无明显变化。MCNP模拟计算表明,对于从热中子至10 MeV的中子,4 cm厚新材料的中子剂量穿透率和中子注量穿透率均优于文献报道的同等厚度的铅硼聚乙烯材料。Am-Be中子源屏蔽试验的实测数据和模拟计算数据表明,两者随屏蔽材料厚度的变化趋势几乎完全一致,两者的差异随屏蔽材料厚度的增加逐渐减小,在10.5 cm处仅1.34%。     

水力驱动控制棒静态特性实验研究

对水力驱动控制步进缸的静态保持特性进行了大量实验研究。获得了水力步进缸的静态保持流量范围和其随温度的变化规律,分析了静态特性与步进缸设计参数之间的关系。结果表明：步进缸的设计参数（重量、对孔尺寸）确定其静态保持流量范围;随温度的升高、最大、最小静态保持流量有所增加;同一温度下不同步位上的静态保持流量范围趋于一致。