SNR-300, strategy and state of the elevated temperature design

This paper describes the procedure for the elevated-temperature structural analysis of the SNR-300. The preliminary analysis based on elastic calculations will be followed by detailed inelastic analysis, if required to provide assurance against all failure modes. In coherence with the progress of the project, experimental work is being done as a first phase of preparation to the elevated temperature design. On the one hand material data will be established and on the other hand analytical work for computer codes and design criteria will be done. The first inelastic investigations are running. Changes in the geometry of the components and/or the planned operating conditions can be considered in case the structural adequacy cannot be demonstrated even through detailed inealstic analysis.     

Numerical analysis of thermal striping induced high cycle thermal fatigue in a mixing tee

Thermal striping, characterized by turbulent mixing of two flow streams of different temperatures that result in temperature fluctuations of coolant near the pipe wall, is one of the main causes of thermal fatigue failure. Coolant temperature oscillations due to thermal striping are on the order of several Hz. Thermal striping high-cycle thermal fatigue that occurs at tee junctions is one of the topics that should be addressed for the life management of light water reactor (LWR) piping systems. This study focuses on numerical analyses of the temperature fluctuations and structural response of coolant piping at a mixing tee. The coolant temperature fluctuations are obtained from Large Eddy Simulations that are validated by experimental data. For the thermal stress fatigue analysis, a model is developed to identify the relative importance of various parameters affecting fatigue-cracking failure. This study shows that the temperature difference between the hot and cold fluids of a tee junction and the enhanced heat transfer coefficient due to turbulent mixing are the dominant factors of thermal fatigue failure of a tee junction.     

A computational fatigue analysis of cyclic thermal shock in notched specimens

Thermal shock induced fatigue plays a role in the assessment of the lifetime of different components in the primary cooling circuit of a nuclear plant. In spite of the implementation of substantial and costly safety factors, a few, unexpected cases of fatigue failure have occurred. Here we report on a laboratory experiment which mimics the thermal loading observed in such components. A finite element thermal stress analysis using a calibrated, elasto-plastic, combined kinematic-isotropic cyclic hardening material model is presented. The distribution of transient stresses and strains in the specimens subjected to cyclic thermal shock, are used to predict the number of cycles to crack initiation with a fatigue curve that has been calibrated experimentally with data from equivalent specimens under pure mechanical fatigue. Our results indicate that cyclic thermal shock induced ratcheting occurs locally near the tip of the notch in the specimens, and the potential impact on the number of cycles to crack initiation is explored.     

Experimental and computational thermal-hydraulic results of fluid and thermal mixing tests

Experimental and computational analyses of a mixing test of cold and hot water flows in a rectangular tee model of the cold leg downcomer geometry of pressurized water reactor were performed. Results obtained from COMMIX-1A computer code calculations showed reasonable agreement with the experimental findings. Counter-current flow and thermal stratification in the cold leg were observed in both the experimental and calculated results for certain ranges of test parameters.     

Comparison of measured and predicted thermal mixing tests using improved finite difference technique

Experimental and computational analyses of a mixing test of cold and hot water flows in a rectangular tee model of the cold leg downcomer geometry of pressurized water reactor were performed. Results obtained from COMMIX-1A computer code calculations showed reasonable agreement with the experimental findings. Counter-current flow and thermal stratification in the cold leg were observed in both the experimental and calculated results for certain ranges of test parameters.     

Study of cleavage initiation under thermal shock by tests on cracked rings and thermomechanical calculations

This article describes a new type of test that easily creates a thermal shock combined with mechanical loading by using standard laboratory equipments. The method consists of rapidly heating a ring initially at −175 °C by circulating hot water through judiciously placed holes. It allows to obtain, at minimal cost, data on crack initiation under thermal shock in the transition zone of steel 16MND5 used for pressure vessels, these data being necessary for work on fracture criteria. Details of two tests are given, one with preloading in which cleavage fracture was observed and one without preloading in which cleavage initiation did not occur. The thermomechanical calculations necessary for the interpretation of both tests are also given. Finally, the global approach to fracture is used involving comparison of the loading paths J(T) with the master curves [ASTM E1921-97, 1997. Standard test method for determination of reference temperature to for ferritic steels in the transition range, ASTM standards], as well as the local approach proposed by Beremin [Beremin, F.M., 1983. A local criterion for cleavage fracture of a nuclear pressure vessel steel. Metall. Trans. A 14A, 2777–2287] for estimating the probability of fracture of such rings. It is shown that, in both approaches, the calculated probabilities are consistent with the experimental observations.     

The effect of paramagnetic shift during thermal quench on internal components in fusion devices

A plasma current disruption is usually initiated by impurity influx that causes a rapid decrease in plasma thermal stored energy (thermal quench). Thermal quench occurs in 500–2000 μs on a large device like ITER. Depending on the β value, the plasma may be either paramagnetic or diamagnetic. Thermal quench causes a large shift in paramagnetism (or diamagnetism) and a corresponding change in toroidal flux. The flux swing can be 1–2 Weber with the rate of change of the toroidal field between 25 and 150 T/s for a device like ITER. The toroidal field shift induces poloidal current in the vessel and possibly in internal components. We have developed a method for simulating the thermal quench field shift that is compatible for use with the electromagnetic simulation codes. The method is based on a radially thin shell having the shape of the last closed flux surface with poloidal current driven to duplicate the toroidal field shift. The magnitude of the current and its time history are adjusted to duplicate the flux change during a disruption thermal quench. We will present the results of using this method to simulate the induced currents in a vacuum vessel having two shells.     

Cumulative damage fatigue tests on nuclear reactor Zircaloy-2 fuel tubes at room temperature and 300°C

Cumulative damage fatigue tests were conducted on the Zircaloy-2 fuel tubes at room temperature and 300°C on the modified Moore type, four-point-loaded, deflection-controlled, rotating bending fatigue testing machine. The cumulative cycle ratio at fracture for the Zircaloy-2 fuel tubes was found to depend on the sequence of loading, stress history, number of cycles of application of the pre-stress and the test temperature. A Hi-Lo type fatigue loading was found to be very much damaging at room temperature and this feature was not observed in the tests at 300°C. Results indicate significant differences in damage interaction and damage propagation under cumulative damage tests at room temperature and at 300°C. Block-loading fatigue tests are suggested as the best method to determine the life-time of Zircaloy-2 fuel tubes under random fatigue loading during their service in the reactor.     

Stress corrosion and thermal fatigue — Experiences and countermeasures in austenitic SS pipings of finnish bwr-plants

A summary of the existence of pipe cracking in Finnish BWR plants is presented covering both thermal fatigue and IGSCC cases. Countermeasures against cracking are evaluated and the measures applied are summarized. Also the results of a research program to monitor ageing of the weld heat affected zones in a pipeline section of a shut-down cooling system are summarized.     

离子束混合中的扩散、热扩散与化学反应效应

离子束混合的理论研究目前集中在混合机制。已提出的机制主要有碰撞级联混合与辐射增强扩散混合。在文献[5、6]中我们给出了这两种机制统一处理的模型与数学表述。     

Irradiation effects on thermal shock resistance and fracture toughness of HTGR-graphite

This paper describes changes in the thermal shock resistance and the thermal shock fracture thoughness in addition to the usual mechanical properties including the diametral compressive strength and fracture toughness of four varieties of graphite for the high temperature gas-cooled reactor due to neutron irradiations of (1.6 2.3) × 10²¹ n/cm² (E > 0.18 MeV) at 600 850°C. These experiments are carried out by using small disk specimens which can be conveniently loaded into a capsule for irradiation in the Japanese Materials Testing Reactor. Both the thermal shock resistance and the thermal shock fracture toughness of graphites after irradiation decreased markedly despite of the increase in mechanical strength.     

Theory and experiment on thermal shock fracture of initially corner-cracked nozzle

The phenomenon of the pressurized thermal shock on the reactor pressure vessel is expected to occur in the case of such an accident as the small loss of the coolant accident in the PWR nuclear plant. In order to study the structural integrity of the reactor pressure vessel under the pressurized thermal shock, the cleavage thermal shock fracture experiment was conducted here using an initially corner-cracked nozzle type specimen made of the pressure vessel steel A508 class 3. The fracture mechanics analysis was performed to asses bthe crack behaviors in the experiment using the time dependent stress intensity factor deduced from the three-dimensional J integral with the thermal effect.     

Low-cycle fatigue studies on nuclear reactor Zircaloy-2 fuel tubes at room temperature, 300 and 350°C

Constant amplitude strain controlled fatigue tests were conducted on the Zircaloy-2 fuel tubes at room temperature, 300 and 350°C on the modified Moore type, four-point-loaded, deflection-controlled, rotating bending fatigue testing machine. The relationship between the plastic strain range, Δ?_p and the number of cyles to failure, N_f was found to be of a simple power law of the form N^β_f · Δ?_p = constant, at all the test temperatures. Strain-hardening coefficients were determined from monotonic tension tests at these temperatures and it was concluded that the material is more or less in a cyclically stable condition.     

Study on mixing behavior in a tee piping and numerical analyses for evaluation of thermal striping

Water experiments were carried out for thermal hydraulic aspects of thermal striping in a mixing tee, which has main to branch diameter ratio of 3. Detailed temperature and velocity fields were measured by a movable thermocouple tree and particle image velocimetry. Flow patterns in the tee were classified into three groups; wall jet, deflecting jet, and impinging jet, which had their own temperature fluctuation profiles, depending on a momentum ratio between the main and branch pipes. Non-dimensional power spectrum density (PSD) of temperature fluctuation showed a unique profile, when the momentum ratio was identical. Numerical simulation based on finite difference method showed alternative vortex development, like Karman vortex series, behind the jet from the branch pipe in the wall jet case. The prominent frequency of the temperature fluctuation in the calculation was 0.2 of St number based on the branch pipe diameter and in good agreement with the experimental results. Mixing behavior in the tee was characterized by the relatively large vortex structures defined by the diameters and the velocities in the pipes.     

The impact of thermal fatigue and other design limits on pulsed commercial tokamak reactor design

Methods of analysis for fusion first-wall design are developed. Several design limits have been evaluated and combined to present tradeoffs in the form of design windows. These considerations include limits related to thermal fatigue, primary membrane strength, displacement under loading, ratcheting, radiation damage, and plasma-wall interactions. Special emphasis is placed on the investigation of thermal fatigue using a two-dimensional treatment of a tubular first-wall configuration. The work is motivated by the proposal of the Ultra Long Pulse Commercial Reactor (ULTR), a machine capable of delivering plasma burn pulses of up to 24 hr in length. The present work looks in detail at the impact of pertinent characteristics of the first-wall design, such as pulse length, coolant pressure, first-wall thickness, and first-wall lifetime on the structural effects considered. Computer programs are developed and are used to consider several major structural effects on a cylindrical first-wall element for both 316 stainless steel and vanadium alloy. Results indicate that short pulse lengths (greater than a few minutes) can be tolerated in tokamak operation. For stainless steel this is true for heat depositions up to 1 MW/m², while vanadium can tolerate heat depositions as high as 2 MW/m². Long pulse operation can be used to increase modestly the allowable heat deposition or to increase useful wall thickness by 1–2 mm. It appears that irradiation swelling and embrittlement, not fatigue, ultimately limits the first-wall design.     

γ辐射灭菌医疗器械初始污染菌的检测、统计与分析

为了解医疗器械生产企业的γ辐射灭菌产品初始污染菌的分布状况，采用IS011737．12006方法对620批次产品进行了检测统计。结果显示，初始污染菌（cfu／件）范围≤1．5的占7．9％、1．5—100（包括100）的占43．5％、100～1000（包括1000）的占26．0％、1000—10000的占18．2％、10000以上的占4．4％，这些结果为γ辐射灭菌的剂量设定和产品质量稳定性提供了有益参考。     

Vacuum plasma-sprayed tungsten on EUROFER and 316L: Results of characterisation and thermal loading tests

Vacuum plasma-spraying (VPS) can be used for the industrial deposition of thick W coatings on actively water-cooled components made of low activation steel or stainless steel. Mock-ups made of martensitic steels, EUROFER and F82H, as well as steel 316L, were coated with 2 mm thick W-VPS layers. The coated materials are candidates for first wall components (ITER and DEMO) receiving moderate heat load of up to 1 MW/m². Mixed tungsten/steel interlayers were applied to reduce the residual and thermal stresses at the substrate–coating interface and to improve the adhesion of the coating. The advantage of this mixed interlayer is that no further (high activation) materials have to be introduced to improve coating adhesion.The characterisation of the W-VPS layers includes the evaluation of the coating microstructure, the measurement of physical and mechanical properties and the metallographical examination before and after heat load tests.Heat load tests with steady state operation up to 2.5 MW/m² and cycling heat loads of 2 MW/m², were successfully completed. They confirm the thermomechanical suitability of industrially manufactured W-VPS coatings for plasma facing first wall components made of steel.     

Homogenisation method for the modal analysis of a nuclear reactor with internal structures modelling and fluid–structure-interaction coupling

A homogenisation method is presented and validated in order to perform the dynamic analysis of a nuclear pressure vessel with a “reduced” numerical model accounting for inertial fluid–structure coupling and describing the geometrical details of the internal structures, periodically embedded within the nuclear reactor. Homogenisation techniques have been widely used in nuclear engineering to model confinement effects in reactor cores or tubes bundles. Application of such techniques to rector internals is investigated in the present paper. The theory bases of the method are first recalled. Adaptation of the homogenisation approach to the case of rector internals is then exposed: it is shown that in such case, confinement effects can de modelled by a suitable modification of classical fluid–structure symmetric formulation. The method is then validated by comparison of 3D and 2D calculations. In the latter, a “reduced” model with homogenised fluid is used, whereas in the former, a full finite element model of the nuclear pressure vessel with internal structures is elaborated. The homogenisation approach is proved to be efficient from the numerical point of view and accurate from the physical point of view. Confinement effects in the industrial case can then be highlighted.     

Preliminary electromagnetic,thermal and structural finite element analysis of FAST load assembly

FAST (Fusion Advanced Studies Torus) is a proposal for a satellite facility of ITER. This current article deals with the development of a complete sequence of finite element models to analyze and verify if the initial geometry chosen for the main structural components of the tokamak called FAST is satisfactory. The first step is concerned with the evaluation of the magnetic field and the following forces produced by the current flowing in the central solenoid, in the outer poloidal coils and in the toroidal ones; the second step is the estimate of the resulting temperatures in the current-carrying conductors; and the third one is the assessment of the state of stress coming from the loads reminded above. The current loads that have been used come from a different analysis that takes account of the equilibrium with plasma. The code employed has been Ansys Rel. 12.