Evaluation of thermal–hydraulic characteristics of reactor coolant system and helically coiled steam generator based on performance tests with SMART-ITL

2021 ◽  
Vol 382 ◽  
pp. 111392
Author(s):  
Hwang Bae ◽  
Sung-Uk Ryu ◽  
Jin-Hwa Yang ◽  
Byong Guk Jeon ◽  
Yoon Gon Bang ◽  
...  
Keyword(s):  
Steam Generator ◽  
Performance Tests ◽  
Hydraulic Characteristics ◽  
Reactor Coolant ◽  
Coolant System

Sensitivity Analysis for Dynamical Response of Reactor Coolant System Based on OPTIMUS

2020 ◽  
Author(s):  
Yuan Yanli ◽  
Ye Xianhui ◽  
Li Lijuan ◽  
Yuan Feng
Keyword(s):  
Sensitivity Analysis ◽  
Steam Generator ◽  
Nuclear Power ◽  
Reactor System ◽  
Response Analysis ◽  
Dynamical Response ◽  
Calculation Error ◽  
Reactor Coolant ◽  
Input Parameters ◽  
Coolant System

Abstract The sensitivity analysis of the dynamical response of reactor coolant system to the input parameters is an important precondition for the design optimization. In this paper, the sensitivity of the dynamical loads at the nozzles of the equipment under seismic conditions is analyzed with an integrated platform called OPTIMUS, taking the stiffness of the dampers in the steam generator and the main pump as the input variables. The key parameters of the reactor system are usually different from the design value due to the calculation error, random and other uncontrollable errors in the manufacturing process and installation process. In a nuclear power project, the measured stiffness values of the dampers on the steam generator and the main pump in the manufacturer are deviated from the requirements in the equipment specification, and it is necessary to evaluate the influence of the deviation on the dynamical response analysis of the reactor system. According to the traditional method, it is necessary to establish the models of the reactor coolant system for nonlinear analysis according to the different stiffness of the dampers, and then the calculation results are compared by EXCEL. In this paper, the sensitivity analysis of output parameters which are the loads at the nozzles of the equipment to the input parameters which are the stiffness of the dampers on the steam generator and pump is realized by OPTIMUS, which is a kind of integration platform. Not only can ANSYS simulation calculations be carried out automatically on the OPTIMUS, but also the output data can be processed rapidly automatically, and the influence of manufacturing deviation of the stiffness of the dampers on the dynamical response of the reactor coolant system can be analyzed quantitatively in the above-mentioned problems, and the data support is provided for the determination of the design variables for subsequent optimization analysis.

Nonlinear Analysis Method Research of Multi-Bolt Connection Structure of Primary Equipment Support

2017 ◽  
Author(s):  
Jianrong Zou ◽  
Shaochong Zhou
Keyword(s):  
Finite Element Method ◽  
Finite Element ◽  
Steam Generator ◽  
Support System ◽  
Analysis Method ◽  
The Finite Element Method ◽  
Reactor Coolant ◽  
Asme Code ◽  
Bolt Connection ◽  
Coolant System

The main equipments of the reactor coolant system include the steam generator, the reactor coolant pump, the pressurizer and the reactor coolant loop. The reactor coolant system is equipped with a steam generator for each of the two loops, and pressurizer is connected with the hot leg of loop 1 using the surge line. The main loop support system design of AP series greatly simplifies the RCS loop support system. Pressurizer supports consist of columns, lower lateral supports, upper lateral support and ring girder, and the steam generator supports consist of columns, lower lateral supports, upper lateral supports and intermediate lateral supports. Ring Girder of pressurizer consists of two semi-circular girders, vertical supports and splice connection of girder and the two half-ring girders are connected with splice connection using 11 bolts. The steam generator upper lateral support is mainly composed of bracket, snubber, pin and ubar and the ubar and the steam generator is connected via 16 bolts in the initial design. These bolts are to ensure the support junction can withstand the force and torque of various conditions of the reactor coolant system, which are important components of the main equipment support. There are large numbers of bolts in the splice connection of ring girder and ubar of upper lateral support of steam generator, and the bolts load was calculated using the uniform method in the general engineering design and analysis. During the design review it was found that the bolts load was uneven and in order to determine the non-uniformity of the bolts the finite element method was used to calculate the load on each bolt, and the resulting stress ratio was greater than 1 did not meet the requirements of the ASME Code. In this paper, the calculation method was studied and the design improvements for parts was made using the nonlinear analysis method to meet the requirements of ASME Code in the case of master supports of main equipment supports have been made good. At the same time the impact of bolts load because of gap was studied. It had very good economic benefits. The calculation and research of this paper show that the finite element method can calculate the force of bolts finely, and can get a more reasonable result than the empirical formula. It can be referred to when the multi-bolt connection structure needs to be refined, such as flange connection of important equipments and valves and flange and bolt design optimization.

scholarly journals FATIGUE LIFE ASSESSMENT OF REACTOR COOLANT SYSTEM COMPONENTS BY USING TRANSFER FUNCTIONS OF INTEGRATED FE MODEL

2010 ◽  
Vol 42 (5) ◽  
pp. 590-599 ◽  
Author(s):  
Shin-Beom Choi ◽  
Yoon-Suk Chang ◽  
Jae-Boong Choi ◽  
Young-Jin Kim ◽  
Myung-Jo Jhung ◽  
...  
Keyword(s):  
Fatigue Life ◽  
Transfer Functions ◽  
Life Assessment ◽  
Fe Model ◽  
Reactor Coolant ◽  
Fatigue Life Assessment ◽  
System Components ◽  
Coolant System

Experimental study on transport behavior of cesium iodide in the reactor coolant system under LWR severe accident conditions

2020 ◽  
Vol 57 (12) ◽  
pp. 1287-1296
Author(s):  
Naoya Miyahara ◽  
Shuhei Miwa ◽  
Mélany Gouëllo ◽  
Junpei Imoto ◽  
Naoki Horiguchi ◽  
...  
Keyword(s):  
Experimental Study ◽  
Cesium Iodide ◽  
Severe Accident ◽  
Reactor Coolant ◽  
Transport Behavior ◽  
Accident Conditions ◽  
Coolant System

Numerical calculation of the thermal-hydraulic characteristics in a once-through steam generator based on a coupled boundary

2019 ◽  
Vol 148 ◽  
pp. 734-746 ◽  
Author(s):  
Jianxin Shi ◽  
Longbin Yang ◽  
Baozhi Sun ◽  
Yanjun Li ◽  
Xiang Yu ◽  
...  
Keyword(s):  
Numerical Calculation ◽  
Steam Generator ◽  
Hydraulic Characteristics
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