Small-Scaled Experimental Research on the Buckling of Steel Containment Under Axial Compression

Focusing on the general and localized elastoplastic buckling of the cylindrical section of steel containment under axial pressure, nonlinear finite element method (FEM) and small-scaled experiments are applied to analysis. First, FEM analysis is conducted considering nonlinear items caused by geometric shape imperfection and elastoplastic constitutive model by the arc-length method RIKS procedure. Parameter sensitivity of the buckling is revealed. Then, small-scaled experiments are carried out. Buckles status is observed, and key geometrical parameters’ influence are found. The results show that cylindrical buckling under axial pressure is sensitive to geometrical parameters and imperfection. It is necessary to employ more realistic parameters to the FEM analysis via accurate geometrical measurement. This research has reference value for the design and fabrication of AP series steel containment vessel.

Technology of Protection Against the Dynamic Effect of Nuclear Pipe Rupture

The special safety facilities and important equipment, etc. of the nuclear power plant will be damaged due to the whipping nuclear high-energy piping after the rupture, and more serious further damage will be caused. In this paper, the process and method of protection analysis of the nuclear high-energy piping rupture have been given from four aspects. The four aspects are location of high-energy piping break, the jet thrust, whipping behavior analysis, and protection analysis of whipping. On the basis of the traditional energy balance method, the method is improved by considering the energy absorbed by the plastic hinge of the piping and the change in the direction of the jet thrust. And then, the comparisons among the traditional energy balance method, the improved energy balance method, and the 3-D finite element dynamic method have been carried out. The deformation of the whip limiter analyzed by the traditional energy balance method is 20.31% larger than which analyzed by the improved energy balance method, and the deformation of the whip limiter analyzed by the 3-D finite element dynamic method is 30.59% smaller than which analyzed by the improved energy balance method. For the first time, a 3-D finite element model according to the true arrangement of the pipe and the whip limiter model are built to simulate the process of whipping not in the plane, considering the energy dissipation of the whip limiter. For the pipe whipping not in the plane and protecting against the pipe rupture by whip limiter, there is no good way to carry out the protection analysis of the piping rupture in the past. Now, the problem can be solved by the 3-D finite element dynamic method.

RPV Structure Integrity Assessment With Surface Cracks Under PTS

As one of the most critical barrier of pressurized-water reactor, Reactor Pressurized Vessel (RPV) is exposed to high temperature, high pressure and irradiation. During the lifetime of RPV, the core belt material will become brittle under the influence of neutron irradiation. The ductile-brittle transition temperature will increase and upper shelf energy will decrease. Thus the structure integrity evaluation of RPV concerning brittle fracture is one of the most important tasks of RPV lifetime management. The non-LOCA accident of Rancho Seco nuclear power plant in 1978 indicates that the emergent cooling transients the sudden cooling down may accompany with the re-pressurize of main loop. The combination of pressure loads and thermal loads may induce a large tensile stress in RPV internal surface, which is the so called pressurized thermal shock (PTS). Due to the existence of welding cladding on the inner surface of RPV, the discontinuity of stress distribution on the cladding-base interface of RPV wall will make calculation of stress-intensity-factor (SIF) difficult. In present research, a two dimensional axial-symmetrical model is built and Finite Element Method (FEM) is adopted to calculate the transient thermal distribution and stress distribution. The influence function method is adopted to calculate crack SIF. Stress distributions in the base and cladding are decomposed respectively and SIFs are calculated respectively to obtain the crack SIF. ASME method is used to calculate the fracture toughness. Present PTS program is validated by the comparative benchmark calculation (the International Comparative Assessment Study of Pressurized Thermal-Shock in Reactor Pressure Vessels). The calculated SIF from present program lies in the reasonable region of the comparing group results. A LOCA transient is investigated with a semi-elliptical surface crack on the RPV beltline region. The temperature and stress distribution along the vessel wall during the transient are given. The stress intensity factors at the deepest and interface point are given respectively. The integrity of RPV under PTS transient is evaluated by comparing stress intensity factor with fracture toughness. Results indicate that the stress intensity factor will not exceed the fracture toughness of the RPV material. The difference between the stress intensity factor and fracture toughness reach a minimum value at the crack tip temperature 20°C. Present research gives a reliable and efficient program to perform RPV structure integrity assessment with surface crack under PTS, which is suitable for further parameter analysis and probabilistic analysis.

Stress Corrosion Behaviors of Three Typical Materials (304L, 2205, Alloy 825) in Boiling Magnesium Chloride Solution

According to YB/T 5362-2006 “stainless steel stress corrosion test method in boiling magnesium chloride solution”, the sensitivity of the stress corrosion of three typical materials (304L, 2205, Alloy 825) was investigated in boiling magnesium chloride solution (experimental temperature is 143±1 °C, concentration of magnesium chloride is 43%). The results show that under the condition of constant strain, the corrosion resistant performance of 825 material is far better than 304L, and the corrosion resistance of dual phase steel may not be superior than that of austenitic stainless steel.

A Study for Evaluating Local Damage to RC Panels Subjected to Oblique Impact: Part 1 — A Study for Evaluating Local Damage Caused by Oblique Impact of Rigid Projectiles

Many empirical formulae have been proposed to evaluate the local damage to reinforced concrete structures caused by the impact of rigid projectiles. Most of these formulae have been derived based on impact tests perpendicular to the target structures. To date, few impact tests oblique to the target structures have been conducted. The purpose of this study is to propose a new formula for evaluating the local damage caused by oblique impacts based on experiments and simulations. The new formula is derived by modifying an empirical formulation for normal impact and the agreement with results of past oblique impact tests is discussed.

Theoretical Analysis of Step-Down of Hydraulic Cylinder Motion for Control Rod Hydraulic Drive System Based on Model II

Hydraulic Control Rod Drive Technology (HCRDT) is a newly invented patent and Institute of Nuclear and New Energy Technology Tsinghua University own HCRDT’s independent intellectual property rights. The hydraulic cylinder is the key part of this technology, so the performance of the hydraulic cylinder directly affects the HCRDT. Firstly, the theoretical model of the cylinder hydraulic has been obtained and verified by the experiment. Second, the step-down process of the cylinder hydraulic is analyzed. The results are shown that the model can analyze the performance of the cylinder, including the motion time of the cylinder, the transient pressure of the cylinder arrival, the transient impact energy of the cylinder arrival. At last, the cylinder and the drive mechanism can be optimized based on the result.

A Study for Evaluating Local Damage to Reinforced Concrete Panels Subjected to Oblique Impact: Part 3 — Simulation Analysis for Evaluating Perforation Phenomena Caused by Oblique Impact of Deformable Projectiles

The purpose of this study is to propose a new formula for evaluating the local damage to reinforced concrete structures caused by oblique impact based on past experimental results and simulation results. In this paper, we present the results of simulation analyses for evaluating the perforation of concrete panels due to oblique impact by deformable projectiles. Various response characteristics and perforation mechanisms such as detailed perforation behavior, damage to reinforced concrete panels, rupture states of the deformable projectiles, reduction in projectile velocity, and residual velocity of the projectile after perforation, and energy transfer processes are clarified. Especially, it is found that the sliding energy consumed by friction occurring at the contact surface between a projectile and a reinforced concrete panel due to oblique impact is remarkably larger than that due to normal impact.

Investigation of Thermal Fatigue Characteristics of Charging Line of RCV in PWR

Nuclear Power Plant (NPP) components need to tolerate thermal constraints, internal pressure and thermal transients. These thermal transients being repeated again and again can lead to thermal fatigue of the component. It has significant effect on the degradation of the NPP components in long term. Studies of thermal fatigue on different NPP components such as mixing tees and valves have been carried out before but the charging line in the chemical and volume control system (RCV) of the NPP seems to have been ignored for thermal fatigue analysis. Charging Line is the connection from RCV towards Reactor Coolant System (RCP). To enhance the safety of the charging line, thermal fatigue evaluation of piping system was performed using the Fluid Structure Interaction (FSI) analysis. Temperature distributions in the pipes were determined via thermal hydraulic analysis (CFX) and the results were applied to the structural model of the piping system to determine the thermal stress (Transient Structural). Results revealed the location of fatigue cracks. Types of stress were identified that caused the fatigue damage. The CFD analysis enabled us to clarify the role of turbulence with respect to the thermal loading of the structure. The study will provide valuable information for establishing a permanent methodology to help minimize thermal fatigue damage in NPP components.

Research and Application on Visual Aided Cable Design System

Nuclear power plants have many types of equipment, dense structure, large plant area, and very complex cable channels. Considering the division and type of cables, cable design has been one of the most complicated tasks in power plant design. Shanghai Nuclear Engineering Research and Design Institute (SNERDI) uses Intergraph PDS system for power plant design, for the AP1000 and CAP series of projects, including equipment and raceway modeling, and use Shaw Cable Manager (SCM) system for cable laying. Cable design work involves multi-system data processing, though without a data integration platform based on 3D visualization. In this paper, a visual aided cable design system is built to display the design data and cable data in a unified 3D plant model, which realizes the visualization and integration of multi-system information, and improves reliability and efficiency of the design work. This method can be applied to a variety of design systems, and has good scalability.

Simulation of Behaviors of a One-Third Scale Containment Model Test

Reactor containment of a nuclear power plant is a structure to ensure the safety of nuclear power plant. It acts as the last barrier to prevent the release of radioactive materials from NPP during accidents. Finite element models were established to simulate a 1/3 scale model of a reactor containment building under leakage test pressure. General finite element software ANSYS were applied. The nonlinear behavior of containment materials, geometric were taken into account in the analysis. The reliability of the finite element model was verified through the comparison of theoretical analysis results with experimental results. In the ANSYS finite element model, the concrete, steel bars and prestress tendons were separated and the prestress tendons were considered by the method of cooling method on the prestress tendon elements. The mechanical properties of the finite element model in the prestress tension process and the absolute internal pressure of 0.52MPa were analyzed. Transient and time dependent losses were taken into account at the same time during the calculation of prestress of tendons, so as to calculate effective prestress at different locations of tendons. Calculation results of prestress losses show that the prestress losses at the hole of equipment hatch are larger than the other areas. The results show that, the deformation of over-all structure of the containment is shrink inward under the action of prestress. And the simulation can achieve the consistent deformation effect between tendons and concrete. The maximum radial displacement of the whole containment structure is located at of 10 ° ∼ 20 °area on the right of the hole of the gate. The effect of expansion deformation of the containment caused by design internal pressure is insufficient to offset the inward shrink effect generated by tendons, and the over-all structure of the concrete containment scale model is mainly under compressive stress. The containment test model is still with a large safety margin under the action of design internal pressure. The largest tensile stress is on the up and down areas of the internal sides of the equipment hatch, dome area close to ring beam, and bottom of perimeter wall close to the base slab. There is possibility of cracking on the concrete in limited local zones. This benchmark can provide a reference for engineering design of containment.