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.

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.

The Airflow Analysis for CAP1400 Control Rod Drive Mechanisms Cooling System

The CAP1400 nuclear power plant (NPP) reactor integrated head package (IHP) refers to the assembly of all of the equipment and structures that are either mounted to the reactor closure head or provide services to the reactor head assembly. One of the main functions of the IHP in the reactor is to provide cooling for the control rod drive mechanism (CRDM) magnetic coils. The IHP cooling system is realized by axial fans connected to the vent tubes of the CRDM cooling shroud. Under normal operating conditions, two of four fans are required to be in operation. The IHP cooling system shall meet the requirements of “keeping the coil temperature within the magnetic coil component below 200°C”. This requirement is achieved by ensuring that the average flow velocity around the CRDM coil assembly is above 15 m/s and that the minimum flow velocity at any location on the outer surface of the coil assembly is not less than 9 m/s. The main purpose of this paper is to study the flow characteristics of the IHP cooling system under various operating conditions. The CFD method is used to obtain the flow field and temperature field in the IHP and to support the rationality of IHP design.

Steam Generator Modal Analysis of Nuclear Power Plant

A method for forming a simplified model of steam generator which will be used in reactor coolant loop analysis has been shown here, as well as the modal analysis to this simplified SG model. This modal analysis results and the results of the SG provided by NPP designer are compared together in order to prove the design correctness. The comparison shows that the two are basically consistent.

Design and Implementation of Redundant Sensors in Active Magnetic Bearing

With the advantages of frictionless, no need of lubrication, no pollution, low consuming and long life, active magnetic bearing (AMB) is applied in the primary helium circulator of the High Temperature Gas-Cooled Reactor-Pebble bed Module (HTR-PM), which is under construction in Shidao Bay Nuclear Power Plant. Active magnetic bearing is a typical mechatronic system with interconnection of mechanical and electronic components with the function of picking up signals, processing and producing. Displacement sensor is an important component to pick up signals for stability control, and also the most susceptible part to fail due to variation of air temperature and vibration of high rotation speed. However, rotating system can’t run normally if a single sensor fails in AMB without redundancy design. For security considerations, higher reliability is required in some special equipment, especially in primary helium circulator of HTR-PM. Design and implementation of redundant sensors is an effective method. This paper reviewed the present research of fault diagnosis and redundant control of displacement sensors, simulation of coil’s short-circuit and open-circuit fault was made with MATLAB/SIMULINK. Parameters were optimized for fault diagnostic circuit by Multisim. Based on the high reliability demand, redundancy design was applied both on structure and control system in AMB. Schematic drawing and PCB board were finished by Altium Design, and experiments were carried out. The result showed that if the coils of sensor failed, AMB system could still work normally by switching to the redundant sensors automatically.

Performance Analysis of Eductor Used for Main Control Room Ventilating in Nuclear Power Plant

This paper proposes the experimental research for the performance of the air eductor used in main control room (MCR). The air eductor is used for emergency ventilating in advanced passive pressurized water reactor in accident. The compress air is supplied to the eductor as a power source and the indoor air is suctioned to the eductor. The performance of the eductor is related to the habitability of MCR. The entrainment ratio and the air pressure of discharge side are the main concerned performance. The entrainment ratio is a value that resulted from the compress air flow rate divided by the suction air flow rate. A test system was set up to test the performance of eductor. The experimental results show that the entrainment ratio of rectangle nozzle with compress air pressure 0.76MPa, 0.80MPa and 0.83MPa were 15.02, 15.04 and 15.06, respectively.

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.